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-045, “ALIQUIPPA FORGE TBD REVISION” 

Contract No. 211-2014-58081 

SCA-TR-PR2014-0092, Rev. 0 

Prepared by 

U. Hans Behling, PhD, MPH 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182 

August 2014 

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 predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42 CFR 82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted.
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S. COHEN & ASSOCIATES:

Technical Support for the Advisory Board on

Radiation & Worker Health Review of

NIOSH Dose Reconstruction Program

Document No. 

SCA-TR-PR2014-0045

Effective Date: 

Draft – August 20, 2014

A REVIEW OF NIOSH’S PROGRAM 

EVALUATION REPORT DCAS-PER-045, 

“ALIQUIPPA FORGE TBD REVISION”

Page 2 of 33

Task Manager:

________________________ Date: ___________

U. Hans Behling, PhD, MPH 

Supersedes:

N/A

Project Manager:

________________________ Date: ___________

John Stiver, MS, CHP

Reviewer:

 John Stiver

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 08/20/2014 Initial issue
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TABLE OF CONTENTS 

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

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

2.0 Relevant Background Information Pertaining to Facility Operations, Health Physics 

Practices and Episodic Efforts to Decontaminate Aliquippa Forge .....................................8

3.0 Subtask 1: Identify the Circumstances that Necessitated the Need for 

DCAS-PER-045 .................................................................................................................11

4.0 Subtask 2: Assess NIOSH’s Specific Methods for Corrective Action .............................13

4.1 An Overview of SC&A’s Previous Review of ORAUT-OTIB-0070 ...................13

4.2 A Focused Review of ORAUT-TKBS-0021, Rev. 01, Pertaining to the 

Residual Period ......................................................................................................16

4.2.1 Estimates of Residual External Exposure ..................................................16

4.2.2 Estimates of Residual Internal Exposure ...................................................18

4.2.3 A Critical Assessment of NIOSH’s Methodology for Deriving 

Inhalation Exposures during the Residual Period ......................................21

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

DRs Requiring Re-evaluation of Dose ..............................................................................25

6.0 Subtask 4: Conduct Audits of a Sample Set of DRs Affected by DCAS-PER-043 .........26

7.0 References ..........................................................................................................................27

Attachment #1: Belmore Memo (1949b) ......................................................................................30

Attachment #2 – Zeuto Survey Instrument ....................................................................................32

Attachment #3 – Belmore Memo (1949a) .....................................................................................33
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ABBREVIATIONS AND ACRONYMS 

Advisory Board 

or Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

AWE Atomic Weapons Employer 

BNI Bechtel National, Inc. 

cm2 square centimeter 

d/y day per year 

D&D Decontamination and Decommissioning 

DCF dose conversion factor 

DOE (U.S.) Department of Energy 

dpm/m3

disintegration per minute per cubic meter 

DR dose reconstruction 

FUSRAP Formerly Utilized Sites Remedial Action Program 

HEPA high-efficiency particular air 

μCi/m microcurie per meter 

μCI/mL microcurie per milliliter 

μg/m3 microgram per cubic meter 

m meter 

MED Manhattan Engineering District 

mR/hr milliroentgen per hour 

mrem millirem 

m/s meter per second 

NIOSH National Institute for Occupational Safety and Health 

NRC (U.S.) Nuclear Regulatory Commission 

NYOO (AEC) New York Operations Office 

OCAS Office of Compensation and Support 

ORAUT Oak Ridge Associated Universities Team 

pCi/d picocurie per day 

PEP Program Evaluation Plan 

PER Program Evaluation Report 

POC probability of causation 
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rem roentgen equivalent man 

RF resuspension factor 

SC&A S. Cohen and Associates (SC&A, Inc.)TBD technical basis document 

TIB technical information bulletin 

U uranium 

wd work day 
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1.0 STATEMENT OF PURPOSE 

To support dose reconstruction (DR), 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 tools. In recognition of the fact 

that all of these supporting elements in DR may be subject to revisions, provisions exist for 

evaluating the effect of such programmatic revisions on the outcome of previously completed 

DRs. Such revisions may be prompted by document revisions due to new information, 

misinterpretation of guidance, changes in policy, and/or programmatic improvements. 

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

DRs has been proceduralized in OCAS-PR-008, Preparation of Program Evaluation Reports 

and Program Evaluation Plans (OCAS 2006), Revision 2, dated December 6, 2006. This 

procedure describes the format and methodology to be employed in preparing a Program 

Evaluation Report (PER) and a Program Evaluation Plan (PEP). 

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%. 

During a teleconference by the Advisory Board’s Procedures Review Subcommittee meeting on 

April 16, 2014, SC&A was tasked by the Board to conduct reviews of three PERs. Included 

among the PERs is DCAS-PER-045, Aliquippa Forge TBD Revision. In conducting a PER 

review, SC&A is committed to perform the following five subtasks, each of which is discussed 

in this report: 

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

on DR. Our assessment intends to ensure that the “issue” was 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. 

Subtask 3: Evaluate the PER’s stated approach for identifying the universe of potentially 

affected DRs, and assess the criteria by which a subset of potentially affected DRs was 

selected for re-evaluation. The second step may have important implications in instances 

where the universe of previously denied DRs is very large and, for reasons of practicality, 

NIOSH’s re-evaluation is confined to a subset of DRs that, based on their scientific 
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judgment, have the potential to be significantly affected by the PER. In behalf of 

Subtask 3, 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 total number of DR audits per PER will be made by the Advisory Board.) 

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

with our review conclusions. 
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2.0 RELEVANT BACKGROUND INFORMATION PERTAINING TO 

FACILITY OPERATIONS, HEALTH PHYSICS PRACTICES AND 

EPISODIC EFFORTS TO DECONTAMINATE ALIQUIPPA FORGE 

The following statements are taken verbatim from ORAUT-TKBS-0021, Rev. 01, Basis for the 

Development of an Exposure Matrix for Aliquippa Forge, Pennsylvania, Period of Operation: 

January 1, 1947 through February 28, 1950 (ORAUT 2004). These statements provide a 

summary as well as a timeline of events that relate to SC&A’s review of DCAS-PER-045 

(DCAS 2013) and its primary technical support document ORAUT-OTIB-0070, Rev. 01, Dose 

Reconstruction during Residual Radioactive Periods at Atomic Weapons Employer Facilities

(ORAUT 2012b). 

(1) The Aliquippa Forge radiological source term consisted primarily of 

natural uranium metal, uranium oxides, and natural uranium’s shortlived progeny. Long-lived progeny prevent significant ingrowth past 234U in 

the 238U decay series and beyond 231Th in the 235U decay series. [Emphasis 

added.]

(2) Vulcan Crucible Steel Company operated Aliquippa Forge to produce 

uranium rods for the AEC from billets primarily by rolling. Operations with 

uranium at the forge began when a trial rolling occurred on July 23, 1948. 

The AEC contract for production work was initiated on August 16, 1948, 

and was extended through February 28, 1950 [AEC 1948]. The rolling 

operation ended on March 30, 1949 with decontamination consuming the 

rest of the contract’s term. . . . [Emphasis added.] 

(3) This analysis assumed that the residual contamination period extended from 

March 1, 1950, through December 31, 1987, and from January 1, 1989, to 

December 31, 1992. [Emphasis added.] 

(4) The site [Aliquippa Forge] consisted of about 19 buildings. The majority of 

the AEC work occurred in Building 3, the rolling mill.

(5) Decontamination was completed by Vulcan Crucible in 1950 in accordance 

with then-current AEC guidelines. [Emphasis added.] 

(6) After the war [WWII] . . . The Medical Division of the AEC New York 

Operations Office (NYOO) felt that a "maximum permissible level" was 

unknown and should be based on human data. Therefore, the 50-μg/m3

level was referred to as the "preferred level" (AEC 1949b). 

(7) As of the September 1, 1948 [AEC] visit, there were few health physics 

controls in place.

(8) Air samples were taken during the September 1948 visit; one sample 

(during the third pass in back of the mill) showed an air concentration as 
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high as 1,800 times the preferred level (i.e., 50 μg/m3). Although peak 

values need to be considered, the fact that work tasks and worker locations 

were constantly changing resulted in time-weighted exposures that were 

typically much lower than the peak values. The other air sampling results 

for September 1 and 2, 1948, showed concentrations in the range of 2.6 to 

510 times the preferred level, . . . [Emphasis added.]

(9) A NYOO report of an AEC visit to the Aliquippa Forge on February 15 and 

16, 1949, described time-weighted radioactive dust exposures between 2.7 

and 5,300 times the preferred level depending on the type of job. [Emphasis 

added.] 

(10) The AEC record indicates that there were no rolling operations after March 

1949 and that only cleanup operations were taking place (Kelley 1949). 

[Emphasis added.] 

(11) In July of 1949, a survey was performed to determine forge cleanup 

requirements (Belmore 1949c). Additional AEC assessments and surveys

were made throughout the cleanup process. Although AEC noted that the 

cleanup personnel had no monitoring equipment, the assessments 

concluded that a sufficient job of cleanup had been done (Eisenbud 1950; 

Belmore 1950). [Emphasis added.] 

(12) The Formerly Utilized Sites Remedial Action Program (FUSRAP) began in 

1976. Perry (1993) stated: 

A radiological survey in 1978 identified contamination (primarily 

uranium-238) in and around onsite buildings. The site was designated for 

further remediation under FUSRAP, and the small operation was shut 

down and the building evacuated. [Emphasis added.] 

(13) In August 1983, the Aliquippa Forge site was designated for remedial action 

under FUSRAP (DOE 1996a). In December 1987, storage activities began 

in Building 3. Interim remedial actions were taken from October to 

December 1988 to enable additional restricted use of Building 3 for 

expansion of a small forging operation . . . Controlled areas were 

established to prevent access to contamination (Seay 1988; DOE 1996a). 

[Emphasis added.] 

(14) Final remedial activities occurred from about June 1993 to September 

1994 (Abelquist 1995; DOE 1996a). [Emphasis added.] 
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SC&A’s Comments 

Based on the above-cited statements that were taken from ORAUT-TKBS-0021, Rev. 01 

(ORAUT 2012a), the following conclusions may be drawn: 

 For a substantial fraction of the facility’s rolling operation period, little effort was made 

to limit personnel exposures and contamination levels. Even the most basic health 

physics practices and facility engineering designs/controls were lacking. 

 Correspondingly, air concentrations during rolling operations were orders of magnitude 

above the AEC recommended “preferred level” of 50 μg/m3(70 dpm/m3).

 Following the operational rolling period that ended on March 30, 1949, some “attempt” 

was made by Vulcan Crucible Steel to decontaminate the facility, which was completed 

in 1950. As noted by the AEC, however, “. . . the cleanup personnel had no monitoring 

equipment” [emphasis added]. Thus, there are no survey data that would quantify 

residual airborne/surface contamination levels. 

 Under FUSRAP, the following activities occurred: 

– A radiological survey of the Aliquippa site was conducted in 1978

– Interim remedial actions were taken to decontaminate the facility in 1988

– Final site remedial activities occurred between June 1993 and September 1994 
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3.0 SUBTASK 1: IDENTIFY THE CIRCUMSTANCES THAT 

NECESSITATED THE NEED FOR DCAS-PER-045 

DCAS-PER-045 was prompted by changes introduced to ORAUT-TKBS-0021, Rev. 00 

(ORAUT 2004), which had been issued on December 21, 2004. 

Revision 01 of ORAUT-TKBS-0021 (ORAUT 2012a) was issued on April 26, 2012, and 

included changes that impacted both internal and external doses that assumedly reflect new 

facility-specific information/data that principally pertain to the residual period of Aliquippa 

Forge and a revision to ORAUT-OTIB-0070, Dose Reconstruction during Residual 

Radioactivity Periods at Atomic Weapons Employer Facilities (ORAUT 2012b). (Note: Neither 

Rev. 00 nor Rev. 01 of ORAUT-TKBS-0021 had previously been reviewed by SC&A.) 

In brief, changes in Rev. 01 of ORAUT-TKBS-0021 (ORAUT 2012a) included the elimination 

of assigned occupational medical x-ray dose; revised time periods that define the residual 

contamination period; an internal dose model, which accounts for exposures during 

Decontamination and Decommissioning (D&D) operations in 1988 and 1993/1994 under 

FUSRAP; and additional data capture information pertaining to survey(s). 

As part of our review of DCAS-PER-045, SC&A compared Rev. 00 (ORAUT 2004) to Rev. 01 

(ORAUT 2012a) of ORAUT-TKBS-0021 to identify all revisions and their consistency with 

stated changes identified in Section 2.0 “Issue Evaluation” of DCAS-PER-045.

SC&A’s Comments Pertaining to the Development of PER-045 and Supporting Documents 

DCAS-PER-045 states the following: 

. . . Revision 1 [of the Aliquippa Forge technical basis document (TBD) ORAUTTKBS-0021] revised the dose estimate in the residual period starting 3/1/1950. 

The revision included both internal and external dose and was the result of both 

new data and a revision to ORAUT-OTIB-0070. [Emphasis added.] 

The statement that this was “. . . the result of . . . a revision to ORAUT-OTIB-0070” is 

misleading/incorrect based on the following dates when documents were issued: 

 Rev. 00 of ORAUT-TKBS-0021, December 21, 2004 (ORAUT 2004)

 Rev. 00 of ORAUT-OTIB-0070, March 10, 2008 (ORAUT 2008)

 Rev. 01 of ORAUT-OTIB-0070, March 5, 2012 (ORAUT 2012b)

 Rev. 01 of ORAUT-TKBS-0021, April 26, 2012 (ORAUT 2012a)

Based on these dates, Rev. 00 of ORAUT-TKBS-0021 predates Rev. 00 of ORAUT-OTIB-0070 

by more than 3 years, which implies the following: 

 Rev. 00 of ORAUT-OTIB-0070 played no role for defining estimates of internal and 

external doses in Rev. 00 of ORAUT-TKBS-0021. A review of ORAUT-TKBS-0021, 
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Rev. 00, shows that internal and external dose estimates were based on assumptions and 

methodologies, which had little in common with ORAUT-OTIB-0070. 

 Thus, it was not the revisions that were introduced in Rev. 01 of ORAUT-OTIB-0070, 

but the existence and substitution of guidance contained in Rev. 01 of ORAUT-OTIB0070 for earlier assumptions stated on page 20 of ORAUT-TKBS-0021, Rev. 00. 

Observation #1. NIOSH should rephrase the role of ORAUT-OTIB-0070 in Section 2.0 of 

DCAS-PER-045 

Observation #2. Review of records indicates that neither Rev. 00 nor Rev. 01 of the Aliquippa 

Forge TBD (ORAUT-TKBS-0021) was ever reviewed/audited by SC&A 

Based on a comparison of estimated internal and external doses in Rev. 00 (ORAUT 2004) and 

Rev. 01 (ORAUT 2012a) of ORAUT-TKBS-0021 (see Section 4.0 of this report), SC&A sees no 

value in a formal retroactive review/audit of Rev. 00 of ORAUT-TKBS-0021. However, as part 

of our audit of DCAS-PER-045, there is a need to critically review Rev. 01 of ORAUT-TKBS0021 with regard to the methodology by which exposures were derived during the residual 

period for the Aliquippa Forge facility.
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4.0 SUBTASK 2: ASSESS NIOSH’S SPECIFIC METHODS FOR 

CORRECTIVE ACTION 

In instances where the PER involves a technical issue that is supported by a document that 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. 

4.1 AN OVERVIEW OF SC&A’S PREVIOUS REVIEW OF ORAUT-OTIB-0070 

SC&A reviewed Rev. 00 of ORAUT-OTIB-0070 in August 2008. Our review identified a total 

of 15 findings that included 4 conditional findings. In behalf of these findings, it was SC&A’s 

opinion that surrogate models/data and specific default values recommended in Rev. 00 of 

ORAUT-OTIB-0070 (ORAUT 2008) for the derivation of air concentrations were likely to 

underestimate inhalation doses. Surrogate models of concern included those identified in 

NUREG-1400 and in Attachment B of ORAUT-OTIB-0070 and assumed default values

pertaining to the source term depletion rate of 1% per day and to the resuspension of 

residual contamination of 1 × 10-6 m

-1

.

Revision 01 of ORAUT-OTIB-0070 (issued on March 5, 2012) deleted the NUREG-1400 source 

term approach and Attachment B, as well as revised the source term depletion rate from 0.01 d-1

to 0.00067 d-1. With the exception of the footnote in Table 5-1, the resuspension factor of 

1 x 10-6 m

-1, however, remained unchanged.

In total, Rev. 01 of OTIB-0070 identifies the following six recommended methods that may be 

used to estimate inhalation dose resulting from residual contamination. Selection from among 

the following six methods is based on availability of data involving air sampling and surface 

contamination levels: 

Method #1: Use of air sampling data taken during facility operations and post-facility 

operations allows for the determination of the exponential decline in air concentration 

defined by the surface contamination depletion factor coefficient (i.e., λ). 

Method #2: Use of operational air sampling data that are adjusted by a source term 

depletion factor to determine post-operational air concentrations from residual 

contamination. 

Method #3: Use of exponential data fitting that are based on empirical post-operational 

air sampling data and estimated air activities during the facility’s operational period. 

Method #4: Use of surface contamination data taken during facility operations and postfacility operations. Air concentrations from residual contamination are derived for any 

given year by means of an empirically derived surface contamination source term 

depletion constant (i.e., λ) and an assumed resuspension factor. 

Method #5: Use of surface contamination data taken during the operational period and 

the application of the resuspension factor 1 × 10-6 m-1 provides the basis for estimating 
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air concentrations during the operational period. To estimate air concentrations during 

any post-operational year, the previously derived operational air concentration is reduced 

by the source term depletion factor defined by λ = 0.00067 per day. 

Method #6: Use of surface contamination data taken during the post-operational period 

and the application of the resuspension factor 1 × 10-6 m-1 provides an estimate of the 

post-operational air concentration. For the operational period, a default air concentration 

is assigned, which, in combination with the derived post-operation air concentration, 

permits an exponential fit. 

These six methods are summarized in Table 5-1 of OTIB-0070 (ORAUT 2012b) and reproduced 

herein as Table 1. The choice of method used to derive internal dose is based on the availability 

data, and the sequence of methods cited in Table 1 identifies the order of preference. 

Table 1. Recommended Methods 

Air Sample Surface Contamination

Recommended Methodology Operational PostOperational Operational PostOperational

X X Exponential fit of operational and post-operational data.

X

Calculate annual intake quantities based on a source term 

depletion factor of 0.00067 d-1(Section 3.5).

X

Exponential fit of post-operational data and estimate of 

operational airborne radioactivity based on ORAUT 

(2006) or Battelle (2011).

X X

Conversion of surface activity to airborne concentrations 

using resuspension factor*or 1 × 10-6 m

-1

followed by an 

exponential fit of derived levels.

X

Conversion of surface activity to airborne concentrations 

using resuspension factors. Calculate annual intake 

quantities based on a source term depletion factor of 

0.00067 m-1(Section 3.5).

X

Conversion of post-operational surface activity data to 

airborne concentrations using resuspension factor*of 1 ×

10-6 m

-1

. Estimate of operational airborne radioactivity 

based on ORAUT (2006) or Battelle (2011). 

Exponential fit of two quantities.

*

 In cases where the contaminated area is still involved in active operations, a site-by-site analysis of the 

appropriateness of the 1 × 10-6 m-1 resuspension factor should be done. 

Thus, from among the six recommended methods, SC&A regards Method #1 as likely to yield 

inhalation estimates that most closely correspond to actual inhalation.1This is due to the fact 

that no assumptions/default values are required (e.g., an assumed resuspension factor, an 

assumed source term depletion factor, representative surface contamination levels, building 

ventilation rates, etc.). However, for Method #1 to yield credible results, the following 

conditions must apply: 

(1) Operational air sampling data should reflect representative work locations 

 

1

 SC&A notes that Method #1 was identified as the method of choice for bounding internal exposures from 

residual contamination in behalf of the Dow Chemical Company/Madison Site (NIOSH 2008). 
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(2) Operational air sampling data should correspond to time(s) that approach the end 

of facility operations 

(3) Post-operational air sampling data should not be compromised by facility 

operations that involve the processing of commercial sources of uranium or 

thorium, which are not included in 42 U.S.C. § 7384(n)(4) 

(4) Post-operational air sampling data should not post-date any known 

decommissioning/decontamination efforts that would compromise the 

credibility of these data 

Based on data availability, Method #1 derives airborne contamination from residual 

contamination during the post-operational period by means of representative empirical sitespecific air monitoring data obtained during operations and post-operations. Implicit in this 

method is that the post-operational airborne activity and the operational airborne activity are 

related through time that accounts for the exponential source term depletion factor (i.e., λ), as 

given in the following equation: 

Air ActivityPost Operation = Air ActivityOperation e –λt Eq. 1 

When solving for λ, the source term depletion factor provides estimates of airborne activity 

values for any post-operational period. 

For Method #2, when available data are restricted to air sampling during facility operations or 

end of facility operations, there is a need to derive air concentration(s) in behalf of the postoperational period that accounts for source term depletion. Variables that affect depletion 

include physical dimensions of the facility, ventilation rate, and resuspension of surface 

contamination. 

For a default value of source term depletion, Rev. 01 of OTIB-0070 derived the average value of 

0.00067 d-1 from empirical data representing four Atomic Weapons Employer (AWE) facilities 

as cited in Table 4-1 and calculated corresponding source term depletion factors presented in 

Table 4-2 of ORAUT-OTIB-0070, along with the recommendation that “. . . This average 

depletion rate should be used for facilities without specific data.” 

For convenience to the reader, Table 4-2 of ORAUT-OTIB-0070, Rev. 01 is enclosed herein as 

Table 2. 
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Table 2. Adjustment Factors to Account for Depletion of Source Term 

during the Residual Period 

Year Factor

1 1.00E+00

2 7.83E-01

3 6.13E-01

4 4.80E-01

5 3.76E-01

6 2.94E-01

7 2.31E-01

8 1.81E-01

9 1.41E-01

10 1.11E-02

11 8.67E-02

12 6.79E-02

13 5.32E-02

14 4.16E-02

15 3.26E-02

16 2.55E-02

17 2.00E-02

18 1.56E-02

19 1.23E-02

20 9.60E-03

21 7.51E-03

22 5.88E-03

23 4.61E-03

24 3.61E-03

25 2.83E-03

26 2.21E-03

27 1.73E-03

28 1.36E-03

29 1.06E-03

30 on 8.32E-04

4.2 A FOCUSED REVIEW OF ORAUT-TKBS-0021, REV. 01, PERTAINING TO 

THE RESIDUAL PERIOD 

4.2.1 Estimates of Residual External Exposure 

To reconstruct external exposure to residual radioactivity, NIOSH states that: 

To reconstruct external exposure to residual radioactivity, the maximum reported 

exposure rate of 0.014 mR/hr (Adams and Payne 1992a) was back-extrapolated

using the source term depletion rate calculated from the internal data [i.e., 1.15 × 

10-4/d or 0.042/yr]. . . . [and] by assuming that workers were exposed for 

2,000 hr/yr. [Emphasis added.] 

Thus, for the year 1992, Table 5-1 of ORAUT-TKBS-0021 identifies the annual dose of 

0.0028 rem (0.014 mR/hr × 2,000 hr/y × 0.001 rem/mrem = 0.028 rem). By extrapolating the 
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annual 1992 dose of 0.028 rem to the start of the residual period in 1950, Table 5-1 identifies the 

annual dose of 0.157 rem (A1950 = A1992 (e6.042/y)(41 y)) = (0.028 rem)(5.6) = 0.157 rem). 

SC&A’s Comments/Findings 

While SC&A was able to reproduce the residual external penetrating doses cited in Table 5-1 of 

ORAUT-TKBS-0021, these values may have been compromised by two independent errors, one 

of which would have overestimated doses, while the other would have underestimated the 

dose, as explained in Findings #1 and #2. 

Finding #1. Failure to account for a previous D&D Effort. 

As previously noted in Section 2.0 above and in Section 5.0 of ORAUT-TKBS-0021, Rev. 01: 

The Aliquippa Forge site was included in the DOE FUSRAP in August 1983. In 

December 1987, Bechtel National, Inc. (BNI) surveyed Aliquippa Forge for the 

purpose of allowing the use of portions of Building 3 for storage. Interim 

remedial activities were conducted by BNI in 1988 by removing contaminated 

materials and equipment and placing a barricade around the remaining 

contaminated area. 

. . . The 1988 effort was limited in Building 3 and occurred in November and 

December of 1988. Vacuums were fitted with high-efficiency particular air 

(HEPA) filters to clean the floors and walls. Contaminated bricks and soil were 

removed as necessary. [Emphasis added.] 

It is therefore reasonable to conclude that the 1988 “interim remedial activities” (in and around 

Buildings 3 and 8; see Finding #3 below) could have significantly reduced the observed dose rate 

of 0.014 mR/hr taken in 1992. Thus, any backward extrapolation to years prior to 1988 (i.e., 

years 1950 through 1987) would have underestimated external doses. 

Finding #2. Backward extrapolation by means of the NIOSH-derived source term 

depletion factor is inappropriate due to impacts associated with the “interim remedial 

activities” of 1988. 

The NIOSH-derived source term depletion factor of 1.15 × 10-4

d

-1 principally reflects 

removable surface contamination that is subject to repeated resuspension and removal by 

building ventilation and other removal mechanisms that contribute to source term depletion. 

For most surface contaminations, however, it is fixed contamination that overwhelmingly 

represents total (fixed and removable) contamination, as suggested by the following statements 

cited in Section 5.0 of ORAUT-TKBS-0021, Rev. 01: 

In 1992 and 1993, areas in and adjacent to Buildings 3 and 8 were further 

characterized (Abelquist 1994; Adams and Payne 1992a,b). The maximum 

reported exposure rate at 1 m was 0.014 mR/hr (Adams and Payne 1992a). The 

maximum removable surface contamination was 350 dpm alpha/100 cm2

. 
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The decontamination techniques in 1993 and 1994 were much more aggressive 

than in 1988. In addition to HEPA vacuuming, which was the main method in 

1988, mechanical shot blasting, concrete saws, and jack-hammering were 

employed (DOE 1996b). The maximum air concentration during the 1993 to 

1994 cleanup was 1.1 × 10-10 μCi/mL. [Emphasis added.] 

The above-cited value of 1.1 × 10-10 μCi/m is equivalent to 110 dpm/m3. Thus, SC&A 

concludes (1) the use of the NIOSH-derived source term depletion factor principally reflects the 

minor or removable portion of the total contamination; and (2) the empirically obtained 1992 

dose rate of 0.014 mR/hr principally reflects the larger fixed contamination component, which is 

not likely to have significantly changed since the beginning of the residual period in 1950. This 

implies that the use of the source term depletion rate overestimated external doses for all years 

prior to 1992.

4.2.2 Estimates of Residual Internal Exposure 

For reconstruction of internal exposures to residual contamination, NIOSH provided the 

following information in Section 5.0 of ORAUT-TKBS-0021, Rev. 01: 

After the end of AEC rolling operations, a July 1949 survey was performed. 

The survey indicated that the maximum air dust concentration, taken during 

normal operations in the Furnace area, was 5.9 μg/m3or 8.94 dpm/m3

(assuming a specific activity of 1.516 dpm/μg for natural uranium) (Belmore 

1949b). 

To calculate internal exposure from residual activity the analysis assumed that 

all buildings had an air concentration of 8.94 dpm/m3 in 1950. This operational 

air concentration was assumed to have occurred for 1 year with no cleanup. An 

indoor deposition velocity of 0.00075 m/s was applied to calculate a 2.11 × 

105 dpm/m2surface contamination level at the end of operations/start of the 

residual period. A resuspension factor of 1 × 10-6 m-1 was applied to the surface 

contamination level, resulting in an air concentration of 0.211 dpm/m3. A source 

term depletion rate was calculated based on a starting air concentration in 1950 

and the air concentration calculated based on the 1992 survey data (ORAUT 

2012b). The 1992 calculated air concentration of 0.035 dpm/m3 was based on 

applying a resuspension factor of 1 × 10-6m

-1 (Abu-Eid et al. 2002) to the 

maximum removable surface contamination of 350 dpm alpha/100 cm2. Using 

these two air concentrations, a source term depletion rate of 1.15 × 10-4 d

-1

was 

calculated. The ingestion intake rates were calculated using the method 

described in Section 3.0. The estimated daily inhalation and ingestion intake 

rates to residual radioactivity from AEC operations at the site (Table 5-1), were 

calculated by assuming that workers were exposed for 2,000 hr/yr. [Emphasis 

added.] 
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As a convenience to the reader, Table 5-1 in ORAUT-TKBS-0021, Rev. 01, is reproduced below 

as Table 3. Inspection of Table 3 shows the following inhalation rates: 

Year Inhalation (pCi/d)

1950 0.627

1992 0.112

Table 3. Annual Internal and External Exposure to Residual Radioactivity

Year Inhalationa

(pCi U/d)

Ingestionb

(pCi U/d)

Penetratingc

(rem)

Nonpenetratingc

(rem)

1950d0.627 0.029 0.157 0.784 

1951 0.627 0.029 0.157 0.784 

1952 0.601 0.028 0.150 0.751 

1953 0.576 0.027 0.144 0.720 

1954 0.552 0.026 0.138 0.691 

1955 0.530 0.024 0.132 0.662 

1956 0.508 0.023 0.127 0.635 

1957 0.487 0.023 0.122 0.609 

1958 0.467 0.022 0.117 0.584 

1959 0.448 0.021 0.112 0.560 

1960 0.429 0.020 0.107 0.537 

1961 0.412 0.019 0.103 0.515 

1962 0.395 0.018 0.099 0.494 

1963 0.379 0.018 0.095 0.473 

1964 0.363 0.017 0.091 0.454 

1965 0.348 0.016 0.087 0.435 

1966 0.334 0.015 0.083 0.417 

1967 0.320 0.015 0.080 0.400 

1968 0.307 0.014 0.077 0.384 

1969 0.294 0.014 0.074 0.368 

1970 0.282 0.013 0.071 0.353 

1971 0.270 0.013 0.068 0.338 

1972 0.259 0.012 0.065 0.324 

1973 0.249 0.012 0.062 0.311 

1974 0.238 0.011 0.060 0.298 

1975 0.229 0.011 0.057 0.286 

1976 0.219 0.010 0.055 0.274 

1977 0.210 0.010 0.053 0.263 

1978 0.202 0.009 0.050 0.252 

1979 0.193 0.009 0.048 0.242 

1980 0.185 0.009 0.046 0.232 

1981 0.178 0.008 0.044 0.222 

1982 0.170 0.008 0.043 0.213 

1983 0.163 0.008 0.041 0.204 

1984 0.157 0.007 0.039 0.196 

1985 0.150 0.007 0.038 0.188 

1986 0.144 0.007 0.036 0.180 

1987 0.138 0.006 0.035 0.173 

1988e0.132 0.006 0.033 0.166 

1989 0.127 0.006 0.032 0.159 

1990 0.122 0.006 0.030 0.152 

1991 0.117 0.005 0.029 0.146 

1992 0.112 0.005 0.028 0.140 
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Table 3. Annual Internal and External Exposure to Residual Radioactivity

Year Inhalationa

(pCi U/d)

Ingestionb

(pCi U/d)

Penetratingc

(rem)

Nonpenetratingc

(rem)

1993e0.107 0.005 0.027 0.134 

1994e0.103 0.005 0.026 0.129 

1995 0.099 0.005 0.025 0.123 

a. Absorption Types M and S are possible. 

b. Choose same f1-value as used for inhalation per NIOSH (2004). 

c. External doses should be assessed using Exposure (R) DCFs. 

d. The operational period ends on February 28, 1950. Therefore, the residual period is extended from 

1951 back to March 1, 1950. 

e. See text below for additional exposure scenarios from clean-up activities. 

SC&A’s Comments and Findings 

Before addressing the methodology employed for the reconstruction of internal exposure, the 

first step of our audit is to reproduce the values cited in Table 3, using NIOSH’s given data: 

 Verification of Daily Inhalation Rates (1950) 

(1) 1949 Air Concentration: 5.9 μg U/m3= 8.94 dpm/m3= 4.027 pCi/m3

(2) Resultant Surface Contamination (1950) = (4.027 pCi/m3)(0.00075 m/s)(3.1536 × 107s/y) 

 = 9.5246 pCi/m2

(3) Resultant Air Concentration (1950) = (95,246 pCi/m2)(1 × 10-6 m

-1) 

 = 0.0952 pCi/m3

(4) Intake per day (1950) = (0.0952 pCi/m3)(1.2 m3)(8 hr/d)(250 wd/365 d/y) 

 = 0.627 pCi/d 

SC&A’s calculated daily intake of 0.627 pCi/d matches NIOSH’s value of 0.627 pCi/d. 

 Verification of Daily Inhalation Rate (1992) 

(1) Surface Contamination (1992) = 350 dpm100 cm2 = 3.5 × 10

4

dpm/m2 

(2) Air Concentration (1992) = (3.5 × 104dpm/m2) )(1 × 10-6 m

-1) 

 = 0.035 dpm/m3

 = 0.0157 pCi/m3

(3) Intake per day (1992) = (0.0157 pCi/m3)( 1.2 m3)(8 hr/d)(250 wd/365 d/y) 

 = 0.103 pCi/d 

SC&A’s calculated daily intake of 0.103 pCi/d is slightly lower and does not match 

NIOSH’s value of 0.112 pCi/d.
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Verification of Daily Ingestion Intakes 

Section 5.0 of the TBD states that “. . . the ingestion intake rates were calculated using the 

method described in Section 3.0 [of the TBD].” In turn, Section 3.0 states “. . . When inhalation 

intakes are calculated from air concentrations . . . the daily ingestion rate in picocuries can be 

estimated by multiplying the daily air concentration in picocuries per cubic meter by a factor of 

0.2 for an 8-hour workday.”

 For 1950, NIOSH derived an air concentration of 0.211 dpm/m3or 0.095 pCi/m3. When 

multiplied by 0.2 and 250 wd/365 d/y, the daily ingestion of 0.013 pCi/d is obtained. 

 For 1992, NIOSH derived an air concentration of 0.035 dpm/m3or 0.0158 pCi/m3. When 

multiplied by 0.2 and 250 wd/365 d/y, the daily ingestion of 0.00216 pCi/d is obtained. 

Finding #3. Using NIOSH’s approach for deriving inhalation and ingestion rates during 

the residual period, SC&A was unable to match values cited in Table 3 above (Table 5-1 of 

ORAUT-TKBS-0021). 

4.2.3 A Critical Assessment of NIOSH’s Methodology for Deriving Inhalation Exposures 

during the Residual Period 

As stated above, NIOSH’s methodology for deriving inhalation exposures was explained in 

Section 5.0 of ORAUT-TKBS-0021 with the following statements: 

After the end of AEC rolling operations, a July 1949 survey was performed. The 

survey indicated that the maximum air dust concentration, taken during normal 

operations in the Furnace area, was 5.9 μg/m3

or 8.94 dpm/m3(assuming a 

specific activity of 1.516 dpm/μg for natural uranium) (Belmore 1949b). 

To calculate internal exposure from residual activity the analysis assumed that all 

buildings had an air concentration of 8.94 dpm/m3in 1950. This operational air 

concentration was assumed to have occurred for 1 year with no cleanup. An 

indoor deposition velocity of 0.00075 m/s was applied to calculate a 2.11 × 

105

dpm/m2surface contamination level at the end of operations/start of the 

residual period. A resuspension factor of 1 × 10-6 m

-1 was applied to the surface 

contamination level, resulting in an air concentration of 0.211 dpm/m3. 

[Emphasis added.] 

SC&A’s daily ingestion rate of 0.013 pCi/d for 1950 does not match NIOSH’s value 

of 0.029 pCi/d.

SC&A’s daily ingestion rate of 0.00216 pCi/d for 1992 is less than half of NIOSH’s 

value of 0.005 pCi/d.
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Comments Pertaining to the Derivation of Air Concentration. Important to note here is that 

NIOSH identified an air concentration of 8.94 dpm/m3for all buildings at Aliquippa Forge 

facility in 1950. SC&A reviewed the reference for the cited air concentration of 8.94 dpm/m3

in Belmore 1949b, which is enclosed herein as Attachment #1. Highlighted on page 2 of 

Attachment #1 is the referenced air sample value of 5.9 μg/m3

(or 8.94 dpm/m3), which was 

taken in the Furnace Area and characterized by NIOSH as the “maximum air dust 

concentration.” 

Referenced on page 2 of Attachment #1, however, is another air sample that at 119 μg/m3

(or 

180 dpm/m3) was 20-fold higher and characterized by the following statement: 

During floor sweeping of the mill area the sample showed 119 micrograms per 

m

3

, this being the only sample in excess of the preferred level. [Emphasis added.] 

Finding #4. Failure to acknowledge and use a reported air sample that at 180 dpm/m3 was 

~20-fold higher than the cited value of 8.94 dpm/m3, which NIOSH described as “the 

maximum air dust concentration taken during normal operations.” 

In addition to NIOSH’s failure to acknowledge the much higher air sample, there is the 

illogical/inexplicable method by which the empirically obtained air concentration of

8.94 dpm/m3 was converted to a modeled value of 0.211 dpm.m3by means of the two 

unsupported model parameters that include (1) the indoor deposition velocity of 0.00075 m/s 

and (2) a resuspension factor of 1 × 10-6 m

-1 as described in the above-cited excerpt from 

Section 5.0 of ORAUT-TKBS-0021, Rev. 01. 

In summary, NIOSH started with an empirically obtained air concentration of 8.94 dpm/m3

in 

1950, and rather than using this value, NIOSH inexplicably converted this to a “modeled” air 

concentration value that at 0.211 dpm/m3 is 42.4 times lower. 

The 42.4-fold error/discrepancy between the observed air concentration of 8.94 dpm/m3and the 

modeled value of 0.211 dpm/m3

also suggests a serious flaw in the NIOSH model; that is to say 

that if the deposition and resuspension of contaminants were truly known, the modeled value 

would have equaled/approximated the observed value of 8.94 dpm/m3.

SC&A concludes that the 42-fold discrepancy reflects deposition and/or resuspension values that 

are too low and inappropriate for Aliquippa Forge during operational periods that did not involve 

AEC contract work. 

In our review of OTIB-0070, Rev. 0 issued in August 29, 2008 (SC&A 2008), SC&A questioned 

NIOSH’s recommended resuspension value of Finding #13, which is reproduced below. 

Indoor resuspension factors cited in the scientific literature that involve 

substantial industrial activities (as would be expected in an AWE facility that may 

continue to operate after operations were performed for the MED or AEC) may 

experience air concentrations corresponding to RF values of 10-4m

-1 to 10-3

m

-1

. 
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The recommended RF value of 1 × 10-6m

-1 by the NRC is limited to facilities that 

have previously been subjected to extensive decontamination efforts in 

anticipation of license termination. A principal objective of such 

decontamination efforts is to minimize any removable surface contamination that 

would contribute to the resuspension of surface contaminant. NRC’s criteria that 

include the recommended RF value of 10-6m

-1 have only recently been 

established, and have limited relevance to AWE facilities that operated decades 

ago and were never subjected to the stringent D&D criteria that separated the 

AWE’s operational period from its post-operational period. 

Perhaps in response to SC&A’s Finding #13, NIOSH added the following footnote to Table 5-1 

of ORAUT-OTIB-0070, Rev. 01, (enclosed herein as Table 1 above): 

* In cases where the contaminated area is still involved in active operations, a 

site-by-site analysis of the appropriateness of the 1 × 10-6m

-1 resuspension factor 

should be done. 

With regard to post-AEC contract operations at Aliquippa Forge, the following statements were 

cited in Section 1.0 of ORAUT-TKBS-0021, Rev. 00 (but not in Rev. 01 of ORAUT-TKBS0021): 

. . . At the time of U.S. Atomic Energy Commission (AEC) contract operations, 

Aliquippa Forge was known as Vulcan Crucible Steel Company. Vulcan Crucible 

was primarily involved with the rolling of natural uranium. Sometime after AEC 

operations ended, the facility became known as Universal Cyclops. The facility is 

now owned by the Beaver County Corporation for Economic Development. 

Finding #5. NIOSH’s “conversion” of the empirically measured air concentration of 

8.94 dpm/m3that was reduced more than 42-fold to a “modeled air concentration” 

represents a major error as the starting point for deriving internal dose for the inhalation 

and ingestion and for all years from 1950 to 1995. 

Finding #6. Inappropriate use of the resuspension factor 1 × 10-6 m

-1 for post-AEC work, 

but nevertheless active operations at the Aliquippa Forge facility as specified in the 

footnote of Table 5-1 in ORAUT-OTIB-0070. 

In behalf of Finding #6, NIOSH had the obligation to conduct a site-specific analysis of available 

survey data that would have suggested a different resuspension value. Presented below are 

examples of data that could have been used to conduct a site-specific analysis of credible 

resuspension values that comply with ORAUT-OTIB-0070. 

Sample Calculations of Resuspension Factors at the Aliquippa Forge Facility Based on Survey 

Data 

The July 28, 1949, AEC survey data at the Aliquippa Forge facility (Belmore 1949b) support 

SC&A’s contention of a higher resuspension factor. For example, page 1 of Attachment #1 
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identifies alpha activities on the Furnace Area Floor of 5,000 to 10,000 dpm that were taken with 

a Zeuto survey instrument, which has a chamber window of 3” × 4” on bottom (or with an 

effective area of 12 inches2) (see Attachment #2). It was in the furnace area that yielded the 

above-cited air concentration of 5.9 μg/m3

or 8.94 dpm/m3. 

Thus, a Zeuto average measurement of 7,500 dpm for a 12-inch2 window corresponds to 

967,500 dpm/m2. Applying NIOSH’s recommended resuspension factor of 1 × 10-6 would yield 

an air concentration of 0.967 dpm/m3, which is 9.24-fold lower than the measured air 

concentration of 8.94 dpm/m3. Thus, a resuspension factor of 1 × 10-5 would represent a more 

realistic value; and for the observed air concentration of 119 μg/m3or 180 dpm/m3 [that 

corresponded to the “mill floor” contamination levels of about 3,500 dpm during a survey on 

May 9, 1949 (see Attachment #3)], a resuspension factor of about 4 × 10-4 would have applied 

during post-AEC facility operations that included (or was equivalent to) “floor sweeping.” 

Comments Pertaining to NIOSH’s Assessment of the Source Term Depletion Rate. To derive 

internal exposure for years 1951 through 1995, NIOSH used the following approach: 

A source term depletion rate was calculated based on a starting air concentration 

in 1950 and the air concentration calculated based on the 1992 survey data

(ORAUT 2012b). The 1992 calculated air concentration of 0.035 dpm/m3

 was 

based on applying a resuspension factor of 1 × 10-6m

-1 (Abu-Eid et al. 2002) to 

the maximum removable surface contamination of 350 dpm alpha/100 cm2. Using 

these two air concentrations, a source term depletion rate of 1.15 × 10-4 d

-1

was 

calculated. . . . The estimated daily inhalation and ingestion intake rates to 

residual radioactivity from AEC operations at the site (Table 5-1) were calculated 

by assuming that workers were exposed for 2,000 hr/yr. [Emphasis added.] 

A second error in the methodology for estimating internal exposures was the selection of the 

350 dpm/100 cm2removable surface contamination value taken during a 1992 survey. From this 

survey measurement, NIOSH derived two critical values for the estimation of inhalation/ 

ingestion doses: 

(1) The air concentration of 0.035 dpm/m3

for the year 1992; and 

(2) A source term depletion rate of 1.15 × 10-4

d

-1 for deriving internal doses between 1950 and 

1995. 

Finding #7. NIOSH’s choice of the 1992 survey measurement of 350 dpm/100 cm2

removable alpha contamination is compromised by the fact that it post-dates the “interim 

decontamination efforts” conducted from October to December 1988. 
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5.0 SUBTASK 3: EVALUATE THE PER’S STATED APPROACH FOR 

IDENTIFYING THE NUMBER OF DRS REQUIRING REEVALUATION OF DOSE 

Section 3.0 of DCAS-PER-045 identified the set of criteria used to identify 21 DRs that had been 

completed under Rev. 00 of ORAUT-TKBS-0021 and required re-evaluation under the revised 

guidance in Rev. 01 of ORAUT-TKBS-0021. All 21 reworked claims resulted in POCs below 

45%. 

SC&A’s Comments and Findings 

Revisions to ORAUT-TKBS-0021, which mandated DCAS-PER-045, exclusively impacted 

internal and external dose estimates associated with a residual period between 1950 and 1995, 

as summarized in Table 5-1 of the TBD. 

SC&A’s audit of ORAUT-TKBS-0021, Rev. 01, and other pertinent documents not only 

identified significant errors, but more importantly questioned the inexplicable approach used by 

NIOSH to redefine air concentration values for estimates of inhalation and ingestion exposures. 

In combination, errors and flawed methodology essentially invalidated all values cited in 

Table 5-1; contrary to Section 2.0 of DCAS-Per-045, which states that “. . . The revision [to 

ORAUT-TKBS-0021] . . . was the result of . . . a revision to ORAUT-OTIB-0070,” [emphasis 

added], guidance as defined in OTIB-0070 was, in fact, ignored 

As shown in Table 1 of Section 4.1 above, prioritization of approaches recommended in OTIB0070 for internal dose assessment would have given priority to Method #2 with the following 

available data: 

(1) The measured air concentration of 8.94 dpm/m3at the end of the rolling period/ 

beginning of the residual period; and 

(2) The revised source term depletion factor of 0.00067 d-1 and values cited in Table 4-2 of 

ORAUT-OTIB-0070. 

Alternatively, second and third tier options would have included Methods #1 and #3, 

respectively. 

.

Finding #8. NIOSH’s methodology for deriving internal inhalation and ingestion dose does 

not comply with the use of available data and the prioritization of recommended methods 

defined in ORAUT-OTIB-0070, Rev. 01.
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6.0 SUBTASK 4: CONDUCT AUDITS OF A SAMPLE SET OF DRs 

AFFECTED BY DCAS-PER-043

Based on the potential magnitude and scope of Findings identified in this audit, SC&A believes 

that any audit of reworked DRs is premature at this time and must await a full review and 

resolution of said Findings. 
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Health for factual accuracy or applicability within the requirements of 42 CFR 82.

7.0 REFERENCES 

42 CFR 82, 2002. Methods for Radiation Dose Reconstruction under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol.67, No. 85/Thursday, May 2, 2002, p. 22314. 

42 U.S.C. §§ 7384 et seq., Energy Emplooyees Occupational Illness Compensation Program Act 

of 2000; as amended; OCAS website. 

Abelquist, E. W., 1994. Additional Characterization Survey of Buildings 3 and 8, Aliquippa 

Forge Site, West Aliquippa, Pennsylvania, ORISE 94/B-77, Final Report, Oak Ridge Institute for 

Science and Education, Oak Ridge, Tennessee, February. [SRDB Ref ID: 6285] 

Abelquist, E.W. 1995. Verification Survey of Buildings 3 and 8, Aliquippa Forge Site, West 

Aliquippa, Pennsylvania, ORISE 95/G-92, Final Report, Oak Ridge Institute for Science and 

Education, Oak Ridge, Tennessee, July. [SRDB Ref ID: 6313] 

Abu-Eid, R.M., R.B. Codell, N.A. Eisenberg, T.E. Harris, and S. McGuire, 2002. Re-Evaluation 

of the Indoor Resuspension Factor for the Screening Analysis of the Building Occupancy 

Scenario for NRC’s License Termination Rule, NUREG-1720, Draft Report for Comment, U.S. 

Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, 

Washington, DC, June. [SRDB Ref ID: 22400] 

Adams, W.C., and Payne, J.L., 1992a. Radiological Survey of the Aliquippa Forge Site, West 

Aliquippa, Pennsylvania, ORISE 93/A-5, Final Report, Oak Ridge Institute for Science and 

Education, Oak Ridge, Tennessee, December. [SRDB Ref ID: 6309] 

Adams, W. C., and Payne, J. L., 1992b. Characterization Survey of Portions of Aliquippa Forge 

Site West Aliquippa, Pennsylvania, ORISE 93/A-6, Final Report, Oak Ridge Institute for Science 

and Education, Oak Ridge, Tennessee, December. [SRDB Ref ID: 6288] 

AEC (U.S. Atomic Energy Commission) 1948. Contract AT-(30-1)-407, August 16. [SRDB Ref 

ID: 6321, p. 2] 

Battelle, 2011, Site Profiles for Atomic Weapons Employers that Worked Uranium Metals, 

Battelle-TBD-6000, Rev. 1, June 1, 2011. 

Belmore, F.M., 1949a. “April 26 Visit to Vulcan Crucible Steel Co., West Aliquippa, Pa.,” 

memorandum to B.S. Wolf, U.S. Atomic Energy Commission, New York Operations Office, 

May 9. [SRDB Ref ID: 9943] 

Belmore, F.M., 1949b. “Decontamination of Vulcan Crucible Steel Co at West Aliquippa, 

Penna.,” memorandum to M. Eisenbud, U.S. Atomic Energy Commission, July 28. [SRDB Ref 

ID: 9939, p. 9] 
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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Belmore, F.M., 1949c. “Clean-Up and Property Disposal,” letter to J. O. Flower (Vulcan 

Crucible Steel Company), U.S. Atomic Energy Commission, September 12. [SRDB Ref ID: 

9939, p. 2] 

Belmore, F.M., 1950. “Clean Up and Property Disposal,” letter to J. O. Flower (Vulcan Crucible 

Steel Company), U.S. Atomic Energy Commission, January 19. [SRDB Ref ID: 6321, p. 14] 

DCAS 2013. Aliquippa Forge TBD Revision, Rev. 00, DCAS-PER-045, NIOSH’s Division of 

Compensation Analysis and Support, Cincinnati, Ohio. May 13, 2013. 

DOE (U.S. Department of Energy) 1996a. Certification Docket for Remedial Action Performed 

at the Aliquippa Forge Site in Aliquippa, Pennsylvania, Oak Ridge Operations Office, Oak 

Ridge, Tennessee, November. [SRDB Ref ID: 13769] 

DOE (U.S. Department of Energy) 1996b. Post-Remedial Action Report for the Aliquippa Forge 

Site, DOE/OR/21949-384, Oak Ridge Operations Office, Oak Ridge, Tennessee, May. [SRDB 

Ref ID: 14407] 

Eisenbud, M., 1950. “Contamination at Vulcan Following First Cleanup,” memorandum to F.M. 

Belmore, U.S. Atomic Energy Commission, January 12, 1950. [SRDB Ref ID: 6321, p. 13] 

Kelley, W.E. 1949. Monthly Status and Progress Report for April 1949, U.S. Atomic Energy 

Commission, New York Operations Office, New York, New York, May 10, 1949. [SRDB Ref 

ID: 6291, p. 93] 

NIOSH 2004. Estimation of Ingestion Intakes, OCAS-TIB-009, Rev. 0, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, April 13. 

NIOSH 2008. SEC Petition Evaluation Report, Petition SEC-00079, Addendum 2. June 3, 2008. 

National Institute for Occupational Safety and Health: Cincinnati, Ohio. 

NUREG-1400 1993. Air Sampling in the Workplace, Final Report, U.S. Nuclear Regulatory 

Commission, Office of Nuclear Regulatory Research, Washington, D.C., September. [SRDB Ref 

ID: 20129] 

OCAS 2006. Preparation of Program Evaluation Reports and Program Evaluation Plans, 

OCAS-PR-008, Rev. 2, NIOSH’s Office of Compensation Analysis and Support, Cincinnati, 

Ohio. December 6, 2006. 

ORAUT 2004. Basis for the Development of an Exposure Matrix for Aliquippa Forge, 

Pennsylvania, Period of Operation: January 1, 1947 through February 28, Rev. 00, ORAUTTKBS-0021. Oak Ridge Associated Universities Team, Cincinnati, Ohio. December 21, 2004. 

ORAUT 2006. Estimating the Maximum Plausible Dose to Workers at Atomic Weapons 

Employer Facilities, ORAUT-OTIB-0004, Revision 3 PC-2, Oak Ridge Associated Universities 

Team, Cincinnati, Ohio. December 6, 2006. 
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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

ORAUT 2008. Dose Reconstruction during Residual Radioactivity Periods at Atomic Weapons 

Employer Facilities, Rev. 00, ORAUT-OTIB-0070. Oak Ridge Associated Universities Team, 

Cincinnati, Ohio. March 10, 2008. 

ORAUT 2012a. Basis for the Development of an Exposure Matrix for Aliquippa Forge, 

Pennsylvania, Period of Operation: January 1, 1947 through February 28, Rev. 01, ORAUTTKBS-0021. Oak Ridge Associated Universities Team, Cincinnati, Ohio. April 26, 2012. 

ORAUT 2012b. Dose Reconstruction during Residual Radioactivity Periods at Atomic Weapons 

Employer Facilities, Rev. 01, ORAUT-OTIB-0070. Oak Ridge Associated Universities Team, 

Cincinnati, Ohio. March 5, 2012. 

Perry, T.C., 1993. “Aliquippa Forge Site – Notification of Uranium Guidelines to Be Used in 

Uranium Remediation Activities,” letter to J. G. Yusko (Pennsylvania Department of 

Environmental Resources), U.S. Department of Energy, Oak Ridge Field Office, Oak Ridge, 

Tennessee, May 17, 1993. [SRDB Ref ID: 13769, p. 158] 

SC&A 2008. Review of ORAUT-OTIB-0070: Dose Reconstruction during Residual 

Radioactivity Periods at Atomic Weapons Employer Facilities, SCA-TR-TASK3-0009, Rev. 0. 

SC&A, Inc., Vienna, Virginia, August 29, 2008. 

Seay, W. M. 1988, “Additional Information Regarding Interim Cleanup of the Universal Cyclops 

Site,” memorandum to J. Wagoner, U.S. Department of Energy, Oak Ridge Operations Office, 

Oak Ridge, Tennessee, November 10. [SRDB Ref ID: 13769, p. 104] 
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Health for factual accuracy or applicability within the requirements of 42 CFR 82.

ATTACHMENT #1: BELMORE MEMO (1949B) 
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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Attachment #1 – Belmore Memo (1949b) (continued) 
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ATTACHMENT #2 – ZEUTO SURVEY INSTRUMENT 
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ATTACHMENT #3 – BELMORE MEMO (1949A)