Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute of Occupational Safety and Health

Review of the NIOSH Site Profile for the 

Linde Ceramics Plant and Tonawanda Laboratory, 

Tonawanda, New York 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0014 

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

July 14, 2006 

Disclaimer 

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

and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, the ABRWH and 

its contractor, SC&A, caution the reader that at the time of its release, this report is pre-decisional and has not 

been reviewed by the Board for factual accuracy or applicability within the requirements of 42CFR82. This 

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

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

regarding its conclusions are unwarranted.
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Document No. 

 SCA-TR-TASK1-0014 

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S. COHEN & ASSOCIATES: 

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program

Revision No. 0 (Draft) 

REVIEW OF THE NIOSH SITE PROFILE FOR THE 

LINDE CERAMICS PLANT AND TONAWANDA 

LABORATORY, TONAWANDA, NEW YORK 

Page 2 of 126

Task Manager: 

_________________________ Date:______________ 

Joe Fitzgerald 

Project Manager: 

_________________________ Date:______________ 

John Mauro, PhD, CHP 

Supersedes: 

 N/A 
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TABLE OF CONTENTS 

Acronyms and Abbreviations ..........................................................................................................5 

1.0 Executive Summary.............................................................................................................8 

1.1 Summary of Strengths............................................................................................16 

1.2 Summary of Findings.............................................................................................18 

1.3 Opportunities for Improvement .............................................................................21 

2.0 Scope and Introduction ......................................................................................................24 

3.0 Assessment Criteria and Method .......................................................................................27 

3.1 Objectives ..............................................................................................................27 

3.1.1 Objective 1: Completeness of Data Sources.............................................27 

3.1.2 Objective 2: Technical Accuracy..............................................................27 

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

3.1.4 Objective 4: Consistency Among Site Profiles.........................................28 

3.1.5 Objective 5: Regulatory Compliance........................................................28 

3.2 Assessment Methodology ......................................................................................29 

4.0 Site Profile Strengths.........................................................................................................33 

4.1 Completeness of Data Sources...............................................................................33 

4.2 Technical Accuracy/Claimant Favorability ...........................................................34 

4.3 Adequacy of Data ..................................................................................................35 

4.4 Consistency Among Site Profiles..........................................................................35 

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

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

5.1 Discussion of Issues...............................................................................................36 

5.1.1 Unsupported Assumptions and Significant Uncertainties in 

Information Used in this Site Profile .........................................................36 

5.1.2 Occupational Internal Dose........................................................................38 

5.1.3 Occupational External Dose ......................................................................47 

5.1.4 Occupational Environmental Dose ............................................................62 

5.2 Other Observations................................................................................................64 

6.0 Overall Adequacy of the Linde Site Profile as a Basis for Dose Reconstruction..............65 

6.1 Satisfying the Five Objectives...............................................................................65 

6.1.1 Objective 1: Completeness of Data Sources.............................................66 

6.1.2 Objective 2: Technical Accuracy..............................................................68 

6.1.3 Objective 3: Adequacy of Data.................................................................70 

6.1.4 Objective 4: Consistency Among Site Profiles.........................................72 

6.1.5 Objective 5: Regulatory Compliance........................................................76 

6.2 Usability of Site Profile for Intended Purpose.......................................................77 

6.3 Unresolved Policy or Generic Technical Issues ....................................................78 
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7.0 References..........................................................................................................................80 

Attachment 1: Conference Call with NIOSH and SC&A.............................................................83 

Attachment 2: Advance Questions for Site Expert Interviews.....................................................98 

Attachment 3: Summary of Site Expert Interviews....................................................................106 

Attachment 4: Issue Resolution Matrix for Findings and Key Observations.............................122 

 

LIST OF TABLES 

Table 1-1: Issue Matrix for the Linde Site Profile ................................................................. 15 

Table 5-1: Time Periods......................................................................................................... 38 

Table 5-2: Available Film Badge/Survey Data and Surrogated Data Used ........................... 57 

Table 5-3: Complex Surrogate External Dose Scheme*........................................................ 58 

Table 6-1: Issue Matrix for the Linde Site Profile ................................................................. 66 

Table 6-2: Site Profile Key Element Comparison.................................................................. 73 

LIST OF FIGURES 

Figure 1-1: Location and Vicinity of Linde Site...................................................................... 12 

Figure 1-2: Linde Site Map...................................................................................................... 13 
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ACRONYMS AND ABBREVIATIONS

Advisory Board Advisory Board on Radiation and Worker Health 

AEC U.S. Atomic Energy Commission 

α Alpha particle 

AWE Atomic Weapons Employer 

β Beta particle 

CFR Code of Federal Regulations

Ci Curie (3.7 × 1010 disintegrations/sec) 

cm Centimeter 

D&D Decontamination and decommissioning 

DOE U.S. Department of Energy 

dpm Disintegration per minute 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 

2000 

EPA U.S. Environmental Protection Agency 

FUSRAP Formerly Utilized Sites Remedial Action Program 

GSD Geometric standard deviation 

GM Geometric mean 

HCA High contamination area 

HEPA High-efficiency particulate air 

ICRP International Commission on Radiological Protection 

in Inch 

INL Idaho National Laboratory 

IREP Interactive RadioEpidemiologic Program 

KERMA Kinetic energy released in matter 

keV Kilo electron volt (1,000 electron volts) 

kg Kilogram 

kV Kilovolt 

kVp Kilovolt peak 

kW Kilowatt 

lb Pound 

LLD Lower limit of detection 

LOD Limit of detection 

LOOW Lake Ontario Ordnance Work 

mA Milliampere 
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MAC Maximum allowable concentration 

MED Manhattan Engineering District 

MCW Mallinckrodt Chemical Works Plant 

MDA Minimum detectable activity 

MDF Materials Development Facility 

mg Milligram 

mR Milliroentgen 

mrad Millirad 

mrem Millirem 

NIOSH National Institute of Occupational Safety and Health 

NYOO New York Operations Office (of the Atomic Energy Commission) 

OCAS (NIOSH) Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities 

ORISE Oak Ridge Institute for Science and Education 

ORNL Oak Ridge National Laboratory 

PA Posterior-anterior 

PAEC Potential alpha energy concentration 

PAEE Potential alpha energy exposure 

PAS Personal air sampler 

PDF Portable document format 

PFG Photofluorography 

PIC Pocket ionization chamber (i.e., "pencil" dosimeter) 

POC Probability of causation 

R Roentgen 

rad Radiation absorbed dose 

rem Roentgen equivalent man 

rep Roentgen-equivalent-physical 

RU Recycled uranium 

SC&A S. Cohen & Associates, Inc. 

SEC Special exposure cohort 

SRS Savannah River Site 

TBD Technical Basis Document 

TCE Trichloroethylene 

TIB Technical Information Bulletin 

TLD Thermoluminescent dosimeter 
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UO3 “Orange oxide” 

UO2 “Brown oxide” 

UF4 “Green salt” 

WEC William Environmental Company 

WL Working level (special unit for exposure to Rn-222 and its progeny) 

WLM Working level month
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1.0 EXECUTIVE SUMMARY 

Under the Energy Employees Occupational Illness Compensation Program Act (EEOICPA, or 

the Act) and Federal regulations defined in Title 42, Part 82, of the Code of Federal Regulations

(42 CFR Part 82), the Advisory Board on Radiation and Worker Health (Advisory Board) is 

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

Institute of Occupational Safety and Health (NIOSH) and its contractors for dose reconstruction. 

As a contractor to the Advisory Board, S. Cohen & Associates (SC&A) has been charged under 

Task 1 to support the Advisory Board in this effort by independently evaluating the approach 

taken in a select number of NIOSH site profiles, that correspond to specific facilities at which 

energy and atomic weapons employees worked and were exposed to ionizing radiation, to gauge 

their adequacy, completeness, and validity. These evaluations will be used by the Advisory 

Board to advise the Secretary of Health and Human Services on the scientific validity, quality, 

and accuracy of dose reconstruction efforts performed by NIOSH and its contractors. 

This report presents the results of SC&A’s evaluation of ORAUT-TKBS-0025, An Exposure 

Matrix for Linde Ceramics Plant (including Tonawanda Laboratory) (Davidson 2005). This 

document is also commonly called the Linde Site Profile and the site, the Linde Site or Linde. 

Our review of the Linde Site Profile focused largely on the quality of available data that 

characterized the facility and its operations, and the methods prescribed by NIOSH for the use of 

those data in dose reconstruction. This review was conducted in accordance with SC&A

Standard Operating Procedure for Performing Site Profile Reviews (SC&A 2004a). 

The Linde Site Profile, supplementing individual claimant exposure data and information 

gathered in interviews with claimants, supports the performance of individual dose 

reconstructions under the EEOICPA. It contains compilations and analyses of data, such as 

those related to facility operations and processes, radiological source term characterizations, 

chemical and physical forms of the radionuclides, historic workplace conditions and practices, 

incidents involving potential exposures, limits of detection of radiation monitoring methods, and 

direction for assigning internal and external doses to monitored and unmonitored workers. 

In addition, SC&A evaluated and made use of the following Technical Information Bulletins 

(TIBs) and the general NIOSH dose evaluation guidelines: 

• ORAUT-OTIB-0002, Technical Information Bulletin – Maximizing Internal Dose 

Estimates for Certain DOE Complex Claims (Rollins 2004) 

• ORAUT-OTIB-0006, Technical Information Bulletin – Dose Reconstruction from 

Occupationally Related Diagnostic X-Ray Projections (Kathren 2003) 

• ORAUT-OTIB-0007, Technical Information Bulletin – Occupational Dose from Elevated 

Ambient Levels of External Radiation (Strom 2003) 
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• ORAUT-OTIB-0023, Assignment of Missed Neutron Doses Based on Dosimeter Records

(Merwin 2005) 

• OCAS-OTIB-009, Technical Information Bulletin – Estimation of Ingestion Intakes 

(NIOSH 2004) 

• OCAS–IG-001, External Dose Reconstruction Implementation Guideline (Taulbee 2002) 

• OCAS–IG-002, Internal Dose Reconstruction Implementation Guideline (Allen 2002) 

The SC&A review also was informed by a number of outside sources and documents, including 

interviews with groups of former and retired Linde workers and site experts (summarized in 

Attachments 2 and 3); Manhattan Engineering District (MED)/New York Operations Office 

(NYOO) (U.S. Atomic Energy Commission) reports; some documents originally provided to 

NIOSH by retired site workers and Linde union representatives; contract and technical 

documents collected by attorneys in charge of the lawsuits brought by claimants; claimant files, 

documents and information used by NIOSH for the development of the Linde Site Profile; and 

other pertinent correspondences, including documents and information stored in the Oak Ridge 

Associated Universities (ORAU) database. 

The SC&A review also took note of a relevant Special Exposure Cohort (SEC) petition. On 

October 12, 2005 (which is subsequent to the issue date of the site profile), NIOSH issued an 

evaluation report on SEC Petition 00044, submitted on September 29, 2005, covering a class of 

atomic weapons employees at the Linde Ceramics Plant from October 1, 1942, through 

October 31, 1947. In this report, NIOSH concluded that dose reconstruction could not be 

completed because of a lack of sufficient information: 

Members of this class at the Linde Ceramics Plant may have received internal 

and external radiation exposures from the uranium and uranium progeny in the 

ores received and processed at the plant. NIOSH lacks any biological monitoring 

data or sufficient air monitoring information or sufficient process and 

radiological source information to estimate the potential airborne concentrations 

to which the proposed class may have been exposed (i.e., internal exposures). 

In light of the conclusions presented in the NIOSH evaluation report, it is not clear how a dose 

reconstructor should perform dose reconstruction for a claimant who worked at the site from 

October 1, 1942, to October 31, 1947. 

The Linde Site was originally owned by Union Carbide, Linde Air Products Division, in 1936. It 

was also known as the Chandler Plant at one time. While portions of the land at the site were 

previously owned by the Town of Tonawanda, the Excelsior Steel Ball Company, the 

Metropolitan Commercial Corporation, and the Pullman Trolley Land Company, the land was 

not used by any of these owners. It is likely that at some point in the past, the land was farmed. 

Commercial industrial processes were conducted by the Linde Air Products Division of Union 

Carbide prior to MED operations in the 1940s. Union Carbide operations continued after the 
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MED-related activities ceased. In the 1990s, Praxair acquired the property and continued to 

perform commercial industrial processes, focusing primarily on research and development. 

From October 1, 1942, to June 30, 1949, portions of the Linde Site were used for the separation 

of uranium ores. These processing activities, conducted under a MED contract, resulted in 

radioactive contamination (ACE 2000). A radiological survey report by Oak Ridge National 

Laboratory (ORNL) in 1978 states that the “site was used for the separation of uranium dioxide 

from uranium ores and for the conversion of uranium dioxide to uranium tetrafluoride during the 

period of 1940–1948” (ORNL 1978). 

The Linde Site was selected for the MED contract because of its experience in the ceramics 

business, which involved processing uranium to produce salts used to color ceramic glazes. 

Under the MED contract, uranium ores from seven different sources were processed in Linde; 

four African (Belgian Congo) ores (three low-grade pitchblendes and torbernite), and three 

domestic ores (carnotite from Colorado) (BNI 1993). 

The domestic ore tailings sent to Linde Site resulted from commercial processing, conducted 

primarily in the Western United States, to remove vanadium. The vanadium removal process 

resulted in disruption of the customary uranium decay chain composition and the removal of 

radium. For this reason, uranium supplied to Linde had low concentrations of radium compared 

to natural uranium and Th-230 concentrations. The African ores shipped to Linde as 

unprocessed mining ores contained uranium in equilibrium with all of the daughter products in 

its decay chain (e.g., Th-230 and Ra-226). The other constituents of the ores were similar to 

those of the domestic ores. Laboratory and pilot plant studies were conducted at Linde from 

1942 to 1943, and uranium processing began at Linde in 1943 (BNI 1993). From mid-1943 to 

mid-1946, a total of about 28,000 tons of ores were processed (Aerospace 1981). 

Five Linde buildings were involved in MED activities; Building 14 (built by Union Carbide in 

the mid-1930s) and Buildings 30, 31, 37, and 38 (built by MED on land owned by Union 

Carbide) (BNI 1993). Ownership of Buildings 30, 31, 37, and 38 was transferred to Linde when 

the MED contract was terminated (BNI 1993). Building 14 served as a pilot plant for the 

separation process carried out in Building 30. A three-phase process was used to separate 

uranium from the uranium ores and tailings. Phase I (conducted in Building 30) consisted of 

separating U3O8 from the feedstock materials by a series of process steps consisting of acid 

digestion, precipitation, and filtration. The filtrate (liquid remaining from the processing 

operations) from this step was discarded as liquid waste into the injection wells, storm sewers, or 

sanitary sewers, and the filter cake was discarded as solid waste and was ultimately transported 

to Ashland 1 (now Ashland Chemical Company in Tonawanda, New York) for storage. The 

U3O8 from Phase I was processed into uranium dioxide (UO3) in Phase 2 (Building 30). In 

Phase 3 (conducted in Buildings 31 and 38), the uranium dioxide was converted to uranium 

tetrafluoride (UF4). Residues from Phases 2 and 3 were reprocessed (Aerospace 1981). 

Building 37 was also used in the above operations (ORNL 1978, p. 2-3). 

The principal solid waste resulting from Phase I was a solid, gelatinous filter cake consisting of 

impurities remaining after filtration of the uranium carbonate solutions. Phase I also produced 
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insoluble precipitates of the dissolved constituents, which were combined with the tailings. The 

precipitated species include large quantities of silicon dioxide, iron hydroxide, calcium 

hydroxide, calcium carbonate, aluminum hydroxide, lead sulfate, lead vanadate, barium sulfate, 

barium carbonate, magnesium hydroxide, magnesium carbonate, and iron complexes of 

vanadium and phosphorus (Aerospace 1981). 

Between 1943 and 1946, approximately 8,000 tons of filter cake from Phase I processing of 

domestic ores were taken from the temporary tailing pile at Linde and transported to Ashland 1. 

These residues contained approximately 0.54% uranium oxide (86,100 pounds of natural 

uranium), which corresponds to 26.5 curies of natural uranium activity. Because the residues 

from the African ore were relatively high in radium content compared with processed domestic 

ore residues, the African ore supplier required that the African ore residues be stored separately, 

so that the radium could be extracted. Between 1943 and 1946, approximately 18,600 metric 

tons (20,500 tons) of residues were shipped to the former Lake Ontario Ordnance Works 

(LOOW) in Lewiston, New York, where they could be isolated and stored in a secure area 

(Aerospace 1981). The production progress reports also showed that approximately 140 metric 

tons (154 tons) of African ore residues were shipped to Middlesex, New Jersey (Aerospace 

1981). 

Subsequent to the uranium operations, a radiation survey was conducted by the Health and 

Safety Division of the NYOO of the Atomic Energy Commission (AEC) in November of 1952 to 

determine the appropriate disposition of the equipment used in the uranium operations. 

Currently, the Linde Site, comprising about 135 acres located at East Park Drive and Woodward 

Avenue, is owned by Praxair, Inc. The site is bounded on the north and south by other industry 

and small businesses, on the east by the CSX railroad tracks and Niagara Mohawk property and 

easements, and on the west by a park owned by Praxair, which is open to the public. The 

regional and vicinity locations of the Linde Site are shown in Figure 1-1. 

The property contains office buildings, fabrication facilities, warehouse storage areas, material 

laydown areas, and parking lots, as shown in Figure 1-2. Access to the site is controlled by 

Praxair. Approximately 1,400 employees currently work there. 

It has not been possible within the time and resources available and SC&A’s scope for this 

review to examine in detail all aspects of the Linde Site Profile, due to the complexity and long 

history of the Linde Site, lack or scarcity of documentation for the early years of operations, and 

the many changes that have occurred over the decades. SC&A has selected only certain issues 

for detailed consideration and discussion, because they may significantly affect dose 

reconstruction, and, ultimately, determination of a claimant’s petition for compensation. 
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Figure 1-1: Location and Vicinity of Linde Site 

(Source: BNI 1982)
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Figure 1-2: Linde Site Map 

(Source: ACE 2000) 
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Based upon the SC&A Standard Operating Procedure for Performing Site Profile Reviews 

(SC&A 2004a), approved by the Advisory Board on March 18, 2004, the SC&A review has 

identified a number of issues. These issues are sorted into the following five categories: 

 

(1) Completeness of data sources 

(2) Technical accuracy 

(3) Adequacy of data 

(4) Consistency among site profiles 

(5) Regulatory compliance 

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

review, this report discusses the strengths of the site profile, followed by a description of the 

issues identified in this review. The issues were carefully assessed with respect to the five 

review criteria listed above. Eighteen of the issues are designated as “findings,”1 because they 

represent what SC&A believes are deficiencies in the site profile that need to be corrected, and 

which have the potential to substantially impact at least some dose reconstructions. The 

remaining issues are designated as “observations,” which represent areas that SC&A feels the 

site profile could improve. 

Issues 

Findings Observations

These issues, and accompanying characterizations and section numbers where they are discussed 

in the report, are listed in Table 1-1. An issue resolution matrix, Table A-5, for these identified 

issues (findings and observations) is provided in Attachment 4 of this report. In addition, 

Section 1.2 summarizes the findings. 

1

 Note: There are only 11 distinct findings, since some of the 18 are combined. 
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Table 1-1: Issue Matrix for the Linde Site Profile 

Descriptions (a)

Issue 

Classification 

Objective 1: 

Completeness 

Of Data 

Objective 2: 

Technical 

Accuracy 

Objective 3: 

Adequacy of 

Data 

Objective 4: 

Site Profile 

Consistency 

Objective 5: 

Regulatory 

Compliance 

Issue 1: (5.1.1) Unsupported 

Assumptions and Significant 

Uncertainties in Information Used 

Finding (1) X X 

Issue 2: (5.1.2.2) Use of Air 

Concentration Data Finding (2) X X X X 

Issue 3: (5.1.2.3) Urinalysis Data Finding (3) X X X X 

Issue 4: (5.1.2.2) Time-Weighted 

Averages Finding (4) X X 

Issue 5: (5.1.2.4) Breathing Rate Observation X X 

Issue 6: (5.1.2.5) Ingestion Rate Observation X X 

Issue 7: (5.1.2.6) Radon Exposure and 

Concentration Observation X 

Issue 8: (5.1.2.7) Raffinate Trace 

Radionuclides Finding (5) X X 

Issue 9: (5.1.2.8) Assigned Work Hours Finding (6) X X 

Issue 10: (5.1.2.9) Surrogate Air 

Concentration Data Finding (7) X 

Issue 11: (5.1.2.10) Use of Geometric 

Mean Values Finding (11) X X 

Issue 12: (5.1.2.11) Lack of 

Comprehensive Uncertainty Analysis Finding (9) X X X 

Issue 13: (5.1.3) Complex Missed 

External Dose Surrogate System Finding (8) X X X 

Issue 14: (5.1.3.4) Film Badge Data Finding (8) X 

Issue 15: (5.1.3.5) Survey Measurement 

Data Finding (8) X 

Issue 16 (5.1.3.6) Time-Weighted 

Averages Finding (4) X X 

Issue 17: (5.1.3.7) Contaminated Burlap 

Bags Observation X 

Issue 18: (5.1.3.8) Surrogate External 

Exposure Data Finding (7) X 

Issue 19: (5.1.3.9) Assigned Work 

Hours Finding (6) X X 

Issue 20: (5.1.3.10) Use of Geometric 

Mean Values Finding (11) X X 

Issue 21: (5.1.3.11) Lack of 

Comprehensive Uncertainty Analysis Finding (9) X X X 

Issue 22: (5.1.4.1) Outdoor Doses Finding (10) X X 

(a) Report section numbers discussing the issues are given after the issue number. 

An “X” in the table indicates significant shortfalls in meeting the corresponding review 

objectives for the indicated topics in the Linde Site Profile. These shortfalls have been discussed 

either within the text of the findings themselves, or, in many cases, in special sections that 

address one or more of these shortfalls. The first column of the table indicates the primary place 

within the report that addresses each issue. 

There is some redundancy in the report by virtue of the standard format adopted, where a single 

item may be discussed from different perspectives in several different places. For example, the 

SC&A site profile review procedure calls for both a “vertical” assessment for purposes of 
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evaluating specific issues of adequacy and completeness, as well as a “horizontal” assessment of 

how the profile satisfies its intended purpose and scope. 

1.1 SUMMARY OF STRENGTHS 

For the purpose of reconstructing occupationally related doses (from medical, internal, external, 

and environmental sources) based on historical operations, NIOSH compiled many documents 

and data describing the radioactive materials, operations, and processes in the various MEDrelated buildings at Linde Site. The Linde Site Profile generally provides a wealth of useful 

information to aid the dose reconstructors, who have the task of determining individual claimant 

radiation exposures. 

The Linde Site Profile, in addition to providing an overview of the site history (Section 2.0), site 

and plant layout (Section 2.1), facilities, and present status, and a framework in which to 

consider occupational exposures, describes the various steps of the uranium processing and 

separation operations in detail. It also describes the past operations and current cleanup efforts 

of seven buildings at Linde Site, including Buildings A, B, 14, 30, 31, 37, and 38, and also their 

associated uranium processing operations (Sections 2.2 and 2.3). The site profile lists major 

radionuclides of concern at these facilities and identifies potential sources of internal and 

external exposures. This information, the first strength, is very helpful to the dose 

reconstructions. 

The most exceptional strength (the second strength) of the site profile is its thorough evaluation 

of different job categories of Linde workers. In some cases, the job categories are grouped into 

high-, medium-, and low-exposure ranking for the purpose of assigning missed doses. The Linde 

Site Profile presents, in Section 2.5, Personnel, Job Categories and Workhours, a list of different 

job titles, job categories, job duties, job functions, and their associated work hours. This jobrelated information is very useful for estimating missed doses by using co-worker or surrogate 

information, if the specific job function or location of the claimant cannot be ascertained. 

The third strength is its extensive statistical analysis of existing airborne uranium dust 

concentrations, radon concentrations, film badge beta and gamma exposure measurements, and 

beta and gamma radiation measurements. In almost all cases, the site profile estimates geometric 

mean (GM) and geometric standard deviation (GSD) values for all potential missed dose 

categories. Based on these GM and GSD values, job categories are grouped into high-, medium-, 

and low-exposure ranking. 

The fourth strength of the site profile is that it attempts, at least in many instances when working 

location, condition, and duration are unknown, to assign claimant-favorable missed doses for 

workers. For example, it assigns a gamma dose of 5.35 R/y for the whole body to all Linde 

Ceramics Plant personnel, based on the results for the most exposed group of monitored process 

workers (Davidson, Table 24, p. 54). 

The major components of the Linde Occupational Internal Dose (Section 3.0) are provided in 

Table 11 for Linde Ceramics Plant workers (p. 37) and Table 12 for Tonawanda Laboratory 
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workers (p. 38). These two tables present default assumptions that could be used to calculate 

missed worker internal dose for the period from October 1, 1942, to December 31, 1954. Missed 

dose default assumptions are based on inhalation intake estimates of radioactive particulates 

presented in Table 9 (p. 34) and uranium progeny activity fractions provided in Table 5 (p. 31). 

However, these two tables only present missed internal dose for workers from August 1, 1947, to 

December 31, 1954. The tables are easy and convenient for dose reconstructors to use if worker 

intake records are not available or complete, and exposures are not too high. On the other hand, 

these tables do not include all potential sources of missed internal doses for workers; these will 

be discussed in the following sections. The tables present recommended missed dose default 

assumptions for two periods of operations as follows: 

(1) October 1, 1942 to October 31, 1947: The default missed worker intake or exposure 

values are designated as “reserved” in this period. In fact, this period has been 

recommended by NIOSH for SEC exemption for occupational internal exposures for 

Linde workers. 

(2) November 1, 1947 to December 31, 1954: The default values or assumptions for this 

period are based on assumed uranium air concentration presented in Section 3.1 (p. 27) 

and assumed radon breath data provided in Section 3.5 (p. 32). 

For missed Occupational Internal Dose after 1954, the site profile presents default worker intake 

values for three radionuclides of concern (i.e., U-234, Th-230, and Ra-226) in Table 39 (p. 74). 

These default values are based primarily on measurements of indoor airborne uranium 

concentrations and radon daughter concentrations. 

Similarly, the major component of the Linde Occupational External Dose (Section 4.0) is 

provided in Table 36, Summary – Annual External Exposure from AWE Operations, 1942–1954

(p. 64). This table presents default missed dose values for beta, gamma, and neutron radiation 

exposures for workers in various years. Another strength of the site profile is its conservatism in 

estimating the missed neutron doses for workers. 

For missed Occupational External Dose after 1954, the site profile presents the default 

penetrating and non-penetrating values in Table 39 (p. 74). These default values are based 

primarily on survey measurements. 

The Occupational Medical Dose section (Section 5) is descriptive in presenting the requirements, 

but lacks substance. The site profile makes many assumptions in setting the relevant parameters, 

and provides no information on types of chest x-ray equipment, radiographic film, 

photoflurographic film, pelvis/lumbar spine x-ray equipment, or on worker medical records. In 

the end, the site profile instructs dose reconstructors to use the NIOSH TIB ORAUT-OTIB-0006, 

Dose Reconstruction from Occupationally Related Diagnostic X-Ray Projections (Kathren 2003), 

as the basis document for reconstructing occupational medical dose. Overall, SC&A believes 

that the default medical doses in ORAUT-OTIB-0006 are claimant favorable for the purpose of 

dose reconstruction. 
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The site profile does not formally evaluate missed occupational environmental doses. However, 

it does evaluate outdoor airborne radioactivity, radon concentrations, beta, and gamma sources. 

In some cases, NIOSH decided to ignore the estimated doses, due to their comparatively small 

quantities. 

1.2 SUMMARY OF FINDINGS 

Table 1-1 presents a list of 22 issues that were identified during the review process. Eighteen of 

these issues were judged potentially important enough to dose reconstruction to categorize as 

findings, and some of them subsequently were combined to yield 11 distinct findings. Those 

findings are summarized below in the order that SC&A believes is their level of significance to 

the dose reconstructions; references are given to report sections where more detailed discussions 

may be found. Specific issues of concern that were identified in our review, and which may 

affect each of the above-cited categories of dose reconstruction, are summarized briefly below. 

Full explanations of these issues are provided in the main text of the report. 

Finding 1 (Issue 1): Unsupported Assumptions and Significant Uncertainties in 

Information Used (Section 5.1.1) – SC&A identified numerous assumptions or values used in 

missed dose estimations (both internal and external) in the site profile that are either not 

supported or not adequately supported by explanation, available data, technical study, or 

references. Many of these assumptions appear somewhat arbitrarily and without adequate 

technical basis. In some cases, an assumption is made or a value selected from a range of 

estimated values in order to bound a dose parameter that is not entirely justified or explained in 

the document. In other cases, the assumption or value selected is not deemed by SC&A as 

bounding. This is a flaw that could affect the credibility and validity of the assigned missed dose 

estimates in this site profile. 

Finding 2 (Issue 2): Use of Air Concentration Data (Section 5.1.2.2) – The use of only 

airborne uranium dust concentration data (air concentration) in the site profile for missed 

occupational internal dose estimation is not defensible or claimant favorable, because there are 

significant uncertainties regarding the use of air concentration data to estimate worker inhalation 

intakes at uranium processing facilities. Several technical studies, including the 2003 Y-12 study, 

Practical Use of Personal Air Sampling (PAS) Data in the Internal Dosimetry Program at the 

Y-12 National Security Complex (Snapp 2003), and the Nuclear Regulatory Commission’s 

NUREG-1400, Air Sampling in the Workplace (Hickey 1993), demonstrate that using air 

concentration data could lead to underestimating the worker intakes and subsequently the 

internal exposures. The Y-12 study shows as much as a 10-times difference between intakes 

derived from bioassay data and intakes derived from air concentration data. In addition, the 

sensitivity of survey instruments, locations of the air sampling, and air flow studies of the 

buildings are not considered in the Linde Site Profile; these factors would impact the accuracy of 

the air concentration data. Also, the air concentration data used are based on results of random 

grab air samples in general areas and breathing zones, but not on continuous area sampling 

measurements in high-risk or high-dose areas. Therefore, SC&A believes that using air 

concentration data only in the Linde Site Profile can lead to significant uncertainties in worker 

inhalation intakes, and the eventual underestimation of missed internal doses. 
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Finding 3 (Issue 3): Neglect of Urinalysis Data (Section 5.1.2.3) – Using air concentration 

data only, and neglecting urinalysis data for estimating worker inhalation intakes in the site 

profile, is not in full compliance with 42 CFR 82 requirements. There are 17 sets of urinalysis 

data for over 100 uranium workers in the ORAU database for the period between December 16, 

1947, and January 30, 1950. The air concentration data used in the site profile are not complete 

to cover all periods of the Linde operation, and, therefore, deemed inadequate (see Finding 2). In 

spite of these shortcomings, NIOSH still decided to use only these air concentration data for 

occupational internal dose reconstruction. This approach is not in full compliance with the 

required “hierarchy of data” approach stipulated in 42 CFR 82. 

Finding 4 (Issues 4 and 16): Time-Weighted Averages (Sections 5.1.2.2 and 5.1.3.6) – Timeweighted averages of internal and external exposure values contain significant uncertainties and 

frequently fail to capture dose to workers in areas of high uranium dust concentration. Even 

using the maximum time-weighted average, dose values would not represent maximized dose 

values and may have limitations when used for denial of claims; nor do they give claimants the 

benefit of the doubt in the face of uncertainties. Individual doses could be far greater than these 

averages, even when the job description, work location, and work duration are known. 

Procedures for estimating 95th percentile values, for instance, would need to be developed in 

which the claimant is given the benefit of the doubt in the face of significant uncertainties. The 

site profile does not evaluate these potential uncertainties and lacks the required procedures for 

dealing with them. Other issues include uncertainties about the lengths of a workday and a 

workweek, as well as number of overtime hours, all of which must be addressed when 

considering time-weighting (see Finding 6). 

Finding 5 (Issue 8): Raffinate Trace Radionuclides Not Addressed (Uranium Progeny) 

(Section 5.1.2.7) – The dose consequences of raffinate trace radionuclides are not adequately 

addressed in the site profile. Specifically, raffinate contains Ac-227 and Pa-231, which are in the 

U-235 decay chain, as well as Th-230, which is part of the U-238 chain. Inhalation of even 

small quantities of some raffinates, such as filter cake (one of the waste products at Linde Site), 

could result in significant doses to the workers. The issue of potential airborne contamination of 

raffinates must be more carefully assessed. In addition, NIOSH lists Ac-227 and Pa-231 in 

Tables 5, 11, and 12 as radionuclides of concern for internal exposure to Linde workers 

(Davidson 2005, p. 37), but does not include them in Table 39 for worker exposure during the 

cleanup period. NIOSH should address these issues as well. In Tables 11 and 12, inhalation 

intake values are not listed for Th-230, Ac-227, and Pa-231 for the period from 1947 to 1954. 

NIOSH should further evaluate the potential exposure pathways for internal exposure of raffinate 

trace radionuclides, and investigate the relative impact of trace radionuclide intakes to the total 

dose. 

Finding 6 (Issues 9 and 19): Inconsistent Work Hours Assigned (Sections 5.1.2.8 and 

5.1.3.9) – The number of work hours used in calculating occupational internal and external doses 

for workers are inconsistent for different periods of Linde operations and, therefore, not claimant 

favorable. Table 4 (Davidson 2005, p. 24), and many other places, state that employees worked 

longer than 40 hours per week. In fact, workers, in some cases, worked typical workweeks of 

9 hours per day for 6 days a week, and for a total of 50 weeks per year. But, in most instances, 
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NIOSH uses the standard 40-hour workweek assumption for the missed dose estimation. This 

approach is not only inconsistent, but also not claimant favorable. 

Finding 7 (Issues 10 and 18): Inappropriate Data Applications as Surrogates for Missed 

Dose Estimation (Sections 5.1.2.9 and 5.1.3.8) – Using the GM of air concentration data of 

seven Atomic Weapons Employer (AWE) facilities in New York from a 1949 AEC/NYOO 

report (AEC 1949a) as surrogate data to develop Linde site-specific worker inhalation intakes for 

the entire period of Linde Operation from 1942 to 1954 is over-reaching, and may underestimate 

the missed occupational internal doses. This approach is inappropriate because the surrogate 

data are very limited and not representative of the actual Linde operation conditions, where 

ventilation was often poor or non-existent and adequate radiation protection practices had not yet 

been developed in the earlier years. The lack of complete film badge data for the period from 

1942 to 1954 at Linde represents a period for which the potential for unaccounted beta and 

gamma doses is greatest. NIOSH’s use of pre-cleanup survey data for the pre-production period 

from 1942 to 1943, the use of eight solid ore samples data for the period from 1943 to 1946, the 

use of a 1-day survey data in six locations in Building 30 for the period from 1946 to 1947, the 

use of two 1-day pre-cleanup survey data after vacuuming and flushing in Building 30 for 1949, 

and the use of post-decontamination survey data for 1950 is complex, over-reaching, 

inadequately supported, and not claimant favorable. In addition, the use of the 1948 film badge 

data collected during the removal of equipment in Building 30 for assigning both beta and 

gamma doses for the period from 1949 to 1954 is not appropriate, because these data do not 

account for external doses to workers from exposures to contaminated burlap bags, contaminated 

soil, and other contaminated sources during the cleanup activities. 

Finding 8 (Issues 13, 14 and 15): Complex Missed External Dose Surrogate System 

(Sections 5.1.3, 5.1.3.4, and 5.1.3.5 – The Linde Site Profile uses a very complex scheme of 

surrogate data to estimate missed occupational external dose to workers from 1942 to the present. 

NIOSH uses a combination of film badge data, solid sample analysis results, and facility field 

measurements to fill exposure data gaps in order to cover the entire period of the Linde 

operations. However, these surrogate data are limited and, most importantly, not facility- or 

building-specific. The procedure for estimating missed beta and gamma doses is not very easy to 

follow, since, in many cases, there are no clear or sometimes any explanation why an assumption 

is made. Even when a clear explanation is presented, the rationale for the assumption is often 

either not consistent or technically sound. These weaknesses lead SC&A to conclude that 

NIOSH’s external dose assumptions are not claimant favorable to workers. 

Finding 9 (Issues 12 and 21): Lack of Comprehensive Uncertainty Analysis for Assigned 

Missed Doses (Sections 5.1.2.11 and 5.1.3.11 – An assessment of uncertainties, as required by 

OCAS-IG-001 and OCAS-IG-002, has not been adequately developed for (1) air concentration 

and radon measurement data used, in lieu of bioassay data, to assign internal dose and for 

external exposure data, including film badge beta and gamma measurements, and (2) survey 

measurements used to assign external dose. As described in the site profile, “little information 

was available” during the periods of production and non-production. In fact, the site profile uses 

different words to indicate if the information gathered is uncertain, such as “probably,” “likely,” 

and “assumes.” It gives the strong impression that the available data gathered are inaccurate and 
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uncertain. NIOSH should develop a method to determine best estimates and their uncertainties, 

as well as the 95th percentile value of time-weighted values of inhalation intake, radon intake, 

beta dose, gamma dose, and neutron dose, for both internal and external dose calculations.

Finding 10 (Issue 21): Outdoor Doses (Section 5.1.4.1) – The Linde Site Profile does not 

discuss environmental contamination or missed occupational environmental doses in detail, 

although there are some source documents in the ORAU database with site-specific 

environmental contamination information. NIOSH did consider, however, several potential 

pathways of outdoor beta and gamma exposures. It also evaluated outdoor radon concentrations 

and some outdoor air radioactivity data for the period from July 2000 to June 2004 (Davidson 

2005, Table 38, p. 72). NIOSH should evaluate more thoroughly all potential environmental 

exposure pathways to workers, including used burlap bags, waste piles, contaminated soils, and 

the contaminated underground tunnel system, and provide clearer guidance in estimating missed 

occupational environmental or outdoor doses. 

Finding 11 (Issues 11 and 20): Use of Geometric Mean Values (Sections 5.1.2.10 and 

5.1.3.10) – The statistical approach used in the Linde Site Profile may not be claimant favorable. 

In Tables 6, 13, 14, 15, 17, 18, 21, 23, 24, 25, 26, 29, 30, 31, 32, and 33, the site profile lists the 

GM or the GSD values for various assigned default assumptions. First, there are no supporting 

calculations or data to show how these geometrical quantities were calculated. Second, the use 

of geometric means and standard deviations of airborne radon concentrations (for example) as 

default values could be considered claimant neutral and not claimant favorable. Unless there is 

good reason to believe that a given worker was exposed to the full distribution of the measured 

concentrations and could not have experienced protracted exposures to higher than average radon 

concentrations, it may be more appropriate to use the upper 95th percentile as the default 

exposure level. NIOSH’s use of the GSD approach may not address very high, short-term, 

episodic exposures; short-term exposure during incidents; and radon intakes during the 

performance of tasks with a potential for high transient air concentrations. 

1.3 OPPORTUNITIES FOR IMPROVEMENT 

Not withstanding the many positive, helpful features of the Linde Site Profile in providing 

guidance to the dose reconstructors, SC&A identified 13 observations (the first 5 of which 

appear in the issue table (Table 1-1) that represent potential opportunities for improvement. 

These are summarized below: 

Observation 1 (Issue 5): Breathing Rate (Section 5.1.2.4) – NIOSH assumed nonconservative, inconsistent breathing rates and breathing types. The site profile assumed a 

breathing rate of 1.2 m3

/hr. This value implies that workers were primarily involved in light 

exercise during the course of the day, certainly not the case for the workers hauling bags of ore. 

A single value may not be consistent with the working conditions in the facility during the early 

years of operation, and is inconsistent with other NIOSH site profiles, such as for Mallinckrodt, 

Bethlehem Steel, and Y-12. In addition, NIOSH has not considered oro-nasal breathing, which 

produces greater deposition in the lung than nasal breathing. NIOSH should consider a breathing 

rate of 1.7 m3/hour for Linde dose reconstruction. 
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Observation 2 (Issue 5): Oro-Nasal Breathing (Sections 5.1.2.4 and 5.2) – The effect of oronasal breathing is relatively important for the workers who were working as loaders during the 

production years, and sweepers and sandblasters during the cleanup years. NIOSH should 

evaluate the potential missed worker doses in these areas. 

Observation 3 (Issue 6): Ingestion Rate (Section 5.1.2.5) – NIOSH estimates worker ingestion 

rates of uranium by multiplying the air concentrations by a factor of 0.2. This factor of 0.2 is 

based on NIOSH’s OCAS-OTIB-009, Estimation of Ingestion Intakes (NIOSH 2004). SC&A 

believes that this factor may not be claimant favorable. As part of SC&A’s review of the 

Bethlehem Steel site profile and its review of OCAS-OTIB-009, SC&A has expressed concern 

that ingestion doses may not be directly related to airborne concentration of radionuclides 

because surface contamination can occur from spills and the direct deposition of flakes of 

uranium oxide onto surfaces. Hence, NIOSH should consider revising this aspect of the Linde 

Site Profile. 

 

Observation 4 (Issue 8): Radon Exposure and Concentration (Section 5.1.2.6) – The Linde 

Site Profile uses the “lowest indoor concentrations measured at the Ceramics Plant during 

African ore processing” as the upper limit to both indoor and outdoor radon concentrations 

during pre-production and initial production periods, because there was no direct measurement 

(Davidson 2005, p. 32). For the period of African ore processing, the assumed outdoor radon 

concentration, 10 pCi/L, is based on the lower limit of detection. This radon concentration value 

is based on the GM of 13 measurements during that period. For the period of domestic ore 

processing, the radon concentration value, both indoor and outdoor, is assumed to be 10 pCi/L. 

SC&A believes these assumed radon concentration values based on the GM of measurements are 

not claimant favorable and not representative of the actual worker exposure conditions during the 

periods of operation from 1942 to 1954. 

Observation 5 (Issue 17): Contaminated Burlap Bags (Section 5.1.3.7) – The Linde Site 

Profile indicates that there were tens of thousand of used burlap bags stacked up in the storage 

area behind Building 30. In addition, during the site expert interviews, past Linde workers 

described that they had been sitting on these contaminated burlap bags during break and lunch 

time. NIOSH should evaluate beta and gamma doses to workers. 

Other observations not identified in the issue table include the following: 

Observation 6: Resuspension Rate (Section 5.2) – NIOSH should consider an exponential 

model for resuspension intakes that takes into account the gradual increase over time in ore dust 

levels and radon levels at Linde facilities. 

Observation 7: Film Badge Data Gap (Section 5.1.3.4) – NIOSH should improve the use of 

film badge data, because significant gaps exist for time periods when workers were not 

monitored for external or internal exposure. 

Observation 8: Use of Site Expert Input (Section 5.2) – It is critical for NIOSH to conduct 

interviews with former workers and other site experts, and integrate first-hand experience and/or 

association with the Linde Site, so as to provide further insight on job category information, site 
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practices/processes/conditions, management practices, and data integrity. NIOSH should make a 

greater effort to take into account site expert information and to investigate worker accounts. 

First-hand experience and association with the Linde facilities enable site experts and workers to 

provide original perspectives and information concerning site practices and exposure histories 

that may not appear in the official records. 

Observation 9: Missing Worker Medical and Exposure Records (Section 5.2) – NIOSH 

should look into the possibility of many missing dose records in Linde worker files. From 

interviews of retired, past, and current workers, there appear to be many incident reports, 

occurrence reports, contamination reports, and worker uptake reports that were not included in 

the worker records. NIOSH holds the position that the dosimetry records accurately reflect all 

the doses workers received while working at Linde. The suspected incompleteness of the worker 

records is a serious issue, since it may lead to significant underestimation of workers’ radiation 

dose. 

Observation 10: Angular Dependence Correction Factors (Section 5.2) – NIOSH should 

provide angular dependence (anatomic geometry) correction factors for external gamma doses, 

particularly for low-photon energies, where the angular dependence of the sensitivity of the 

dosimeter is most pronounced. These correction factors are used to account for, for example, the 

bias introduced by a dosimeter worn at the neck level and the higher doses received by 

tissues/organs below the waist. 

Observation 11: Examples of Dose Calculation (Section 5.2) – NIOSH should provide an 

example (or examples) in the site profile of a hypothetical dose reconstruction using recorded 

records, missed dose assignment, and dose assignments when dosimeters read zero dose. 

Observation 12: Inconsistencies Exist among the Seven Site Profiles Currently under 

SC&A’s Review (Section 6.1.4) – Whereas, dosimeter adjustment factors are applied to 

recorded external dose at the Savannah River Site (SRS), Idaho, Y-12, Rocky Flats, and Hanford 

to estimate Hp(10) doses, the Mallinckrodt and Bethlehem site profiles do not recommend an 

adjustment to recorded film doses. It also appears that the Linde Site Profile did not consider the 

laboratory, radiological, and environmental uncertainties in the personal dosimetry program. 

Lastly, the Linde Site Profile did not consider the occupational dose from environmental 

exposure, as was done in the Idaho, SRS, Y-12, Rocky Flats, and Hanford site profiles. 

Observation 13: Inappropriate Application of Residual Contamination Data (Section 5.1.2)

– Data used for reconstructing potential missed internal and external doses during years of 

residual contamination are not representative of actual conditions. For example, 1976 Building 

30 air concentration data were used for missed internal dose estimation for the entire residual 

period, and 1982 BNI survey data were used for estimating missed external doses. There are 

significant gaps in these surrogate data that do not address potential worker exposures while 

working in the complex and contaminated underground tunnels and handling contaminated waste 

from cleanup activities.
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2.0 SCOPE AND INTRODUCTION 

When the United States government and its contractors first became interested in 

uranium, Linde Air Products, then a division of Union Carbide and Carbon 

Corporation, operated Tonawanda Laboratory, which had been producing U3O8

that was sold as a coloring agent for ceramics. Because of the great interest in 

obtaining uranium that could be used to create the experimental uranium piles, 

Linde was contracted to develop uranium chemical processes and build a facility 

that could process large amounts of uranium ore. This commissioned facility was 

called the Linde Ceramics Plant (Davidson 2005, p. 10). 

Tonawanda Laboratory and the Linde Ceramics Plant were located on land 

owned by Union Carbide at East Park Drive and Woodward Avenue in 

Tonawanda, New York (ORNL 1978, Fig. 3). The site is near the intersection of 

Riverview Boulevard and Woodward Avenue. It is north of Woodward Avenue, 

east of East Park Drive, and west of the Conrail railroad tracks. Tonawanda site 

buildings involved in MED/AEC work are shown in Figure 1. Tonawanda 

Laboratory occupied Building 14, and the Ceramics Plant used Buildings 30, 31, 

37, 38, and A. Building B contained MED offices. Ownership of the Ceramics 

Plant buildings was transferred to Linde after the site cleanup that began with the 

shutdown of production in 1949. The transfer probably was completed in 1954. 

In the 1990s, the site was acquired by Praxair, Inc. (ACE Buffalo 2003). As of 

this writing (2005), it is owned by Praxair (Davidson 2005, p. 11). 

Under the Energy Employees Occupational Illness Compensation Program Act of 2000 

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

Reconstruction Under the Energy Employees Occupational Illness Compensation Program, of 

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

Health (Advisory Board) is mandated to conduct an independent review of the methods and 

procedures used by the National Institute for Occupational Safety and Health (NIOSH) and its 

contractors for dose reconstruction. As a contractor to the Advisory Board, S. Cohen and 

Associates (SC&A) has been charged to support the Advisory Board in this effort by 

independently evaluating a select number of site profiles that correspond to specific facilities at 

which energy employees worked and were exposed to ionizing radiation. 

This report provides a review of the Linde Site Profile, ORAUT-TKBS-0025, An Exposure 

Matrix for Linde Ceramics Plant (including Tonawanda Laboratory) (Davidson 2005), and 

related documents in order to accomplish the following: 

• Determine the completeness of the information gathered by NIOSH in behalf of the site 

profile, with a view to assessing the profile’s adequacy and accuracy in supporting 

individual dose reconstructions 

• Assess the technical merit of the data/information 

• Assess NIOSH’s use of the data in dose reconstructions 
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SC&A’s review of the Linde Site Profile focuses on the quality and completeness of the data 

characterizing the facility and its operations, and the methods prescribed by NIOSH for the use 

of these data in the dose reconstruction process. The review was conducted in accordance with 

Standard Operating Procedure for Performing Site Profile Reviews (SC&A 2004a). The criteria 

for evaluation include whether the Site Profile provides a basis for scientifically supportable dose 

reconstructions in a manner that is adequate, complete, efficient, and (especially) claimant 

favorable. Specifically, these criteria were viewed from the perspective of whether dose 

reconstructions based on the site profile would support robust compensation decisions. 

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

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

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

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

reconstructions, bridging uncertainties, or correcting technical inaccuracies. 

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

workers were exposed, using detailed individual worker dose records to the greatest extent 

possible, and, where gaps exist, supplementing the dose records with site profile data, which 

present potential exposures pathways at different time periods. The hierarchy of data used for 

developing dose reconstruction methodologies is (1) film badge and urinalysis data, (2) coworker data and workplace monitoring data, and (3) process description information or source 

term data. 

In accordance with directions provided by the Advisory Board and the SC&A site profile review 

procedure, this report is organized into the following sections: 

(1) Executive Summary 

(2) Scope and Introduction 

(3) Assessment Criteria and Method 

(4) Site Profile Strengths 

(5) Vertical Issues 

(6) Overall Adequacy of the Linde Site Profile as a Basis for Dose Reconstruction 

Based on the issues identified in each of these sections, SC&A prepared a list (Tables 1-1 

and 6-1) of “issues” briefly described in the Executive Summary and later in the report; this list 

functions as a convenient roadmap to the issues discussed throughout the report. Issues are 

designated as “findings” if SC&A believes that they represent deficiencies in the site profile that 

need to be corrected and which have the potential to have a substantial impact on at least some 

dose reconstructions. Issues are designated as “observations” if they simply raise questions, 

which, if addressed, would improve the site profile and may possibly reveal deficiencies that will 

need to be addressed in future revisions of the site profile. In this review, SC&A has identified 

22 issues, categorized into 11 findings (five sets of 2 issues and one set of 3 issues are combined), 

and 4 observations (plus another 8 observations not connected with an issue). The site profile, in 

many ways, has done a successful job in addressing a series of difficult technical challenges. In 
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other areas, the site profile exhibits shortcomings that may influence some dose reconstructions 

in a substantial manner. 

Since many of the issues that surface in the report correspond to more than one of the major 

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

profiles, and regulatory compliance), there is a degree of redundancy in the SC&A report, where 

different sections may address the same issue, but from different perspectives. For example, the 

SC&A site profile review procedure calls for both a “vertical” assessment for purposes of 

evaluating specific issues of adequacy and completeness, as well as a “horizontal” assessment of 

how the profile satisfies its intended purpose and scope.
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3.0 ASSESSMENT CRITERIA AND METHOD 

SC&A is charged with evaluating the approach set forth in the site profile used in the individual 

dose reconstruction process. SC&A reviewed the site profile documents with respect to the 

degree to which technically sound judgments or assumptions are employed, and assessed the 

degree to which they fulfill the objectives delineated in SC&A’s review procedure. 

3.1 OBJECTIVES 

Documents are reviewed for their completeness, technical accuracy, adequacy of data, 

consistency with other site profiles, and compliance with the stated objectives, as defined in the 

SC&A Standard Operating Procedure for Performing Site Profile Reviews (SC&A 2004a). 

These objectives are discussed in the following sections: 

3.1.1 Objective 1: Completeness of Data Sources 

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

principal sources of data and information that are essential to the development of the site profile. 

The two elements examined under this objective are (1) determining if the site profile makes 

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

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

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

reports or other site documents for particular processes, and if the TBDs have not taken these 

data into consideration, this would constitute a completeness of data issue. The Oak Ridge 

Associated Universities (ORAU) site profile document database, including the referenced 

sources in the technical basis documents (TBDs), was evaluated to determine the relevance and 

use of the data collected by NIOSH to the development of the site profile. Additionally, SC&A 

evaluated selected records publicly available relating to the Linde site and records provided by 

site experts. 

3.1.2 Objective 2: Technical Accuracy 

SC&A reviewed the site profile with respect to Objective 2, which requires SC&A to perform a 

critical assessment of the methods used in the site profile to develop technically defensible 

guidance or instruction, including evaluating field characterization data, source term data, 

technical reports, standards and guidance documents, and literature related to processes that 

occurred at the Linde Site. The goal of this objective is to first analyze the data according to 

sound scientific principles, and then evaluate this information in the context of compensation. 

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

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

3.1.3 Objective 3: Adequacy of Data 

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

whether the data and guidance presented in the site profile are sufficiently detailed and complete 
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to conduct dose reconstruction, and whether a defensible approach has been developed in the 

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

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

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

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

considered a gap and, thus, an inadequacy in the data. 

3.1.4 Objective 4: Consistency Among Site Profiles 

SC&A reviewed the site profile with respect to Objective 4, which requires SC&A to identify 

common elements within site profiles completed or reviewed to date, as appropriate. In order to 

accomplish this objective, the Linde Site Profile was compared to the Mallinckrodt, Bethlehem, 

Y-12, and other site profiles and TBDs. Both the Mallinckrodt and Y-12 site profiles are 

appropriate for comparison, as the sites had similar uranium processes. This assessment was 

conducted to identify areas of inconsistencies and determine the potential significance of any 

inconsistencies with regard to the dose reconstruction process. 

3.1.5 Objective 5: Regulatory Compliance 

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

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

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

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

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

important to briefly review key elements of the dose reconstruction process, as specified in 42 

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

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

dependence on the availability, completeness, and accuracy/uncertainty of available dose data. 

The first two categories represent “extreme cases,” where, in the first (Category 1), exposures are 

so obviously high as to lead quickly to a probability of causation (POC) of at least 50% and, in 

the second (Category 2), exposures are so obviously low, as to lead quickly to determination of a 

POC of less than 50%. The third category (Category 3) is the most difficult one to assess, as the 

claimant’s exposure falls between the two extremes of the first two categories. 

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

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

value in excess of 50%, and assures compensability to the claimant. Such partial/incomplete 

dose reconstructions with a POC greater than 50% may, in some cases, involve only a limited 

amount of external or internal data. In extreme cases, even a total absence of a positive 

measurement may suffice for an assigned organ dose that results in a POC greater than 50%. For 

this reason, dose reconstructions in behalf of this category may only be marginally affected by 

incomplete/missing data or uncertainty of the measurements. In fact, regulatory guidelines 

recommend the use of a partial/incomplete dose reconstruction, the minimization of dose, and 

the exclusion of uncertainty for reasons of process efficiency, as long as this limited effort 

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

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

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

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

365 days per year; while not logical, the assumption is certainly upper bounding on exposure 

time. The use of such maximized (or upper bound) values, however, is limited to those instances 

where the resultant maximized doses yield POC values below 50%, which are not compensated. 

For this second category, the dose reconstructor needs only to ensure that all potential internal 

and external exposure pathways have been considered. 

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

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

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

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

to compensate a claim. 

Although simplistic in design, to satisfy this type of a dose reconstruction, the site profile must, 

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

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

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

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

assignment. Similarly, in order to maximize internal exposures, the highest plausible air 

concentrations and surface contaminations must be identified. 

Category 3: The most complex and challenging dose reconstruction represents claims where the 

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

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

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

implausible results in a POC greater than 50%, denial is not possible. A more refined estimate 

may be required to support a recommendation either to deny or to compensate the claimant. In 

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

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

“reasonable estimates” of radiation dose to mean the following: 

. . . estimates calculated using a substantial basis of fact and the application of 

science-based, logical assumptions to supplement or interpret the factual basis. 

Claimants will in no case be harmed by any level of uncertainty involved in their 

claims, since assumptions applied by NIOSH will consistently give the benefit of 

the doubt to claimants. 

3.2 ASSESSMENT METHODOLOGY 

In order to assess the degree of compliance with the five objectives described above, SC&A 

reviewed the site profile, pertinent TIBs, and other relevant documents, giving due consideration 

to the three categories of dose reconstructions that the site profile is intended to support 
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(Section 3.1.5). The Linde Site Profile generally provides well-organized and somewhat or 

partially user-friendly information for the dose reconstructor when adequate data are available to 

do that comprehensively. During the course of its review, SC&A was cognizant of the fact that 

the site profile is not required by the EEOICPA or by 42 CFR Part 82, which implements the 

statute. Site profiles were developed by NIOSH as a resource to the dose reconstructors for 

identifying site-specific practices, parameter values, and factors that are relevant to dose 

reconstruction, and which may be used to supplement a claimant’s own employment and 

exposure record. Based on information provided by NIOSH personnel, SC&A understands that 

site profiles are living documents, which are revised, refined, and supplemented with TIBs as 

required, to help dose reconstructors. Site profiles are not intended to be prescriptive or 

necessarily complete in terms of addressing every possible issue that may be relevant to a given 

dose reconstruction. In addition, they are not intended for the “layman,” but for the health 

physics personnel immersed in the review process. The principal documents and data sources 

SC&A examined in the course of its review, and which were most influential in informing 

SC&A’s assessment, are the following: 

• ORAUT-TKBS-0025, An Exposure Matrix for Linde Ceramics Plant (including 

Tonawanda Laboratory) (Davidson 2005) – This document represents the entire Linde 

Site Profile. It consists of six major sections; (1) Introduction, (2) Site Description and 

Operational History, (3) Estimation of Internal Exposure, 1942–1954, (4) Estimation of 

External Exposure, 1942–1954, (5) Occupational Medical Exposure, and (6) Estimation 

of Exposures from Residual Contamination After 1954. 

– Section 1 provides a useful overview and explains the role of other sections in support 

of the dose reconstruction process. Hence, the introduction helps in framing the 

scope of the site profile. 

– Section 2, in addition to providing an overview of the site history (Section 2.0), 

layout, facilities (Section 2.1), source term (Section 2.2), activities (Section 2.3), and 

present status, and a framework in which to consider occupational exposures, 

describes in detail the various steps of the uranium processing and separation 

operations. 

– Section 3 describes the various sources of internal exposure to workers at the Linde 

Site. It also presents default assumptions that are to be used by dose reconstructors to 

estimate missed worker internal dose for the period from October 1, 1942, to 

December 31, 1954. 

– Section 4 describes the various available external exposure data sources for dose 

reconstruction. It also presents default missed dose values for beta, gamma, and 

neutron radiation exposures for workers in various years at the site. 

– Section 5 is descriptive in the requirements of occupational medical monitoring. It 

eventually provides instruction to dose reconstructors to use the appropriate NIOSH 
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Technical Information Bulletin (TIB) as the basis for reconstructing occupational 

medical doses. 

– Section 6 presents NIOSH’s approach in estimating internal and external doses to 

workers as a result of residual contaminants in various Linde facilities and locations 

after 1954. 

• ORAUT-OTIB-0002, Technical Information Bulletin – Maximizing Internal Dose 

Estimates for Certain DOE Complex Claims (Rollins 2004) 

• ORAUT-OTIB-0006, Technical Information Bulletin – Dose Reconstruction from 

Occupationally Related Diagnostic X-Ray Projections (Kathren 2003) 

• ORAUT-OTIB-0007, Technical Information Bulletin – Occupational Dose from Elevated 

Ambient Levels of External Radiation (Strom 2003) 

• ORAUT-OTIB-0023, Assignment of Missed Neutron Doses Based on Dosimeter Records

(Merwin 2005) 

• OCAS-OTIB-009, Technical Information Bulletin – Estimation of Ingestion Intakes 

(NIOSH 2004) 

• OCAS-IG-001, External Dose Reconstruction Implementation Guideline (Taulbee 2002) 

• OCAS-IG-002, Internal Dose Reconstruction Implementation Guideline (Allen 2002) 

In accordance with SC&A’s site profile review procedure, the reviewers performed an initial 

assessment of the site profile, its supporting documentation, and several relevant TIBs. SC&A 

then submitted written questions (Attachment 1) to NIOSH with regard to assumptions and 

methodologies used in the site profile. SC&A expects to discuss the issues raised in a 

teleconference with staff members of NIOSH, ORAU and ORAU subcontractors (note, this has 

not happened as of the date of this assessment due to schedule constraint). NIOSH, ORAU, and 

ORAU subcontractor personnel will subsequently be given the opportunity to comment on the 

SC&A account of the teleconference, and the revised account will be included in a revision of 

this report. 

SC&A conducted site expert interviews to help obtain a comprehensive understanding of the 

radiation protection program, site operations, and environmental contamination that might be 

present in some areas. While it is recognized that peoples’ memories may not be wholly reliable, 

especially when trying to recall information from decades ago that may not have seemed 

significant at the time, the interviews, especially taken in the aggregate, provided much useful 

insight from the perspective of the Linde workers themselves. Attachment 2 presents advance 

questions for the site interviews, and Attachment 3 provides summaries of the interviews 

conducted by SC&A in the Tonawanda area during the course of this review. The site experts 

interviewed include current, former, and retired staff from dosimetry, radiation control, 
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operations, environmental monitoring, maintenance, instrumentation, electrical, mechanical, 

security, engineering, management, and other support organizations. Each summary is an edited 

paraphrase of conversations held with a number of site experts, rather than a verbatim transcript. 

Personnel statements have been grouped into categories to provide a linkage with various 

portions of the Linde Site Profile. References that may identify specific site experts have been 

omitted for privacy reasons. Interviewed individuals were given the opportunity to review the 

interview summary for accuracy and corrections were made where necessary. This is an 

important safeguard against missing key issues or misinterpreting some vital piece of 

information. Most, but not all, of the individuals interviewed by SC&A provided comments on 

the summaries. 

Information provided in the expected teleconference with NIOSH mentioned above will be 

evaluated against the preliminary findings to finalize the vertical issues addressed in the report 

(Section 5). There are three levels of review for this report. First, SC&A team members review 

it internally. Second, SC&A engages an outside consultant, who has not participated in the 

preparation of this document, to review all aspects of this report. The third level, referred to as 

the expanded review cycle, will consist of a review of this draft by the Advisory Board and 

NIOSH. The first two review levels have been completed. 

Upon delivery of this draft to NIOSH and the Advisory Board, the report will be posted on the 

Advisory Board’s web site. The report will then under go an expanded review by a Boarddesignated working group. The working group will address of the issues identified in this report 

and, based on direction provided by the working group and the Board, SC&A and/or NIOSH 

may be directed to perform follow-up investigations that could result in revisions to this report or 

to the site profile, or both. 

Finally, it is important to note that SC&A’s review of the Site Profile and its supporting TIBs is 

not exhaustive. These are large, complex documents, and SC&A used its judgment in selecting 

those issues that we believe would be important with respect to dose reconstruction.
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4.0 SITE PROFILE STRENGTHS 

In developing a site profile, the assumptions used must be fair, consistent, and scientifically 

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

development of the site profile must also consider efficiency in the process of analyzing 

individual exposure histories, so that claims can be processed in a timely manner; this is clearly 

in the best interest of the claimants. With this perspective in mind, SC&A identified a number of 

strengths in the Linde Site Profile. These strengths are described in the following sections. 

4.1 COMPLETENESS OF DATA SOURCES 

The site profile exhibits the following strengths in terms of the completeness of its data sources: 

(1) In an effort to be comprehensive in addressing the range of facilities and processes, 

NIOSH compiled facility-specific information from a number of descriptions and 

historical records. NIOSH drew upon information contained in 153 reports and 

documents cited in the reference section of the site profile. Seven buildings were 

identified as the Linde facilities used for uranium-related processing and operations. A 

concerted effort was made to characterize the principal types and relative importance of 

the various radionuclides that may have contributed to internal and external exposures at 

the various facilities and associated processes over the life of the facilities. SC&A 

considers this compilation to be an important strength of the report. In addition, NIOSH 

compiled an historical timeline (Table 1 of the site profile) for the purpose of identifying 

key dates of events, activities, and operations. 

(2) The site profile thoroughly evaluates different job categories of workers for the entire 

period of operation from 1942 to the present. In some cases, the job categories were 

grouped into high-, medium-, and low-exposure ranking for the purpose of assigning 

missed doses. Section 2.5, Personnel, Job Categories and Workhours, presents the list of 

different job titles, job categories, job duties, job functions, and their associated work 

hours for Linde workers. This job-related information is very useful for dose 

reconstructors to estimate missed doses for the claimant by using co-worker or surrogate 

information if the specific job function or location of the claimant cannot be ascertained. 

(3) For the purpose of developing data needed to reconstruct missed internal doses based on 

historical operations, NIOSH compiled a significant amount of data identifying the 

radioactive materials at the Linde Site, and describing the relevant operations and their 

associated processes. Notwithstanding this achievement, however, there are 

opportunities for improvement in the data sets and guidance for the dose reconstructors 

for reconstructing missed internal exposures, and also for identifying missed internal 

doses due to deficient work practices and inadequate instrumentation. 

(4) In compiling historical data needed to reconstruct missed external (gamma, beta, and 

neutron) doses, NIOSH compiled a significant amount of data identifying potential 

external radiation sources at various Linde buildings, and describing the relevant 
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operations and their associated processes. NIOSH also compiled data related to external 

dosimetry over the site’s entire operating history. From these gathered data, NIOSH 

developed the missed gamma and neutron dose values for different time periods. 

Opportunities for improvement remain, however, in the areas of adding the missed beta 

dose values and identifying high-risk (with respect to exposure) jobs. 

4.2 TECHNICAL ACCURACY/CLAIMANT FAVORABILITY 

The Linde Site Profile exhibits the following strengths, in terms of technical accuracy and 

claimant favorability: 

(1) The site profile performed extensive statistical analysis of existing airborne uranium dust 

concentrations, radon concentrations, film badge beta and gamma exposure 

measurements, and beta and gamma radiation measurements. In almost all cases, the site 

profile estimates GM and GSD values for all potential missed dose categories. Based on 

these GM and GSD values, job categories are grouped into high-, medium-, and lowexposure ranking. 

(2) In many instances, when working location, condition, and duration are unknown, the site 

profile attempts to assign the highest missed doses for workers. For example, it assigns a 

gamma dose of 5.35 R/y for the whole body to all personnel based on the results for the 

most exposed group of monitored process workers (Davidson 2005, Table 24, p. 54). 

(3) In lieu of any neutron measurements or personnel neutron monitoring data, the 

occupational neutron dose section of the site profile has adequately bound the potential 

missed neutron dose rates and annual neutron dose to workers. NIOSH’s approach is 

based primarily on potential alpha-neutron sources during the Linde operation period 

from 1943 to 1946. The assigned neutron doses to workers for different operation 

periods are considered to be claimant favorable. 

(4) The Occupational Medical Dose section (Section 5) is descriptive in presenting the 

requirements, but lacks substance. The site profile uses many assumptions in the 

parameters, but provides no information on actual chest x-ray equipment, radiographic 

film, photoflurographic film, pelvis/lumbar spine x-ray equipment, or worker medical 

records. In the end, the discussion is not particularly relevant, since the site profile 

finally instructs dose reconstructors to use the NIOSH TIB ORAUT-OTIB-0006, Dose 

Reconstruction from Occupationally Related Diagnostic X-Ray Projections, as the basis 

document for reconstructing occupational medical dose for workers. Overall, SC&A 

believes that using the default medical doses in ORAUT-OTIB-0006 is claimant 

favorable for the purpose of Linde dose reconstruction. 
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4.3 ADEQUACY OF DATA 

The Linde Site Profile suffers from not having access to adequate historical information and 

exposure data that were compiled as a part of the Linde operations. In most cases, NIOSH was 

forced to make many assumptions that appear arbitrary and unsupported. 

 

4.4 CONSISTENCY AMONG SITE PROFILES 

Although Linde, Mallinckrodt, and Y-12 missions overlapped to a significant extent, there are 

many differences in the facility designs, processes, and radiological practices. In some cases, 

these differences require site-specific assumptions in dose determinations. 

NIOSH has made a concerted effort to recognize and address site-specific issues in the TBDs. 

With respect to the Interactive RadioEpidemiologic Program (IREP) input parameters, the Linde, 

Mallinkrodt, and Y-12 site profiles are consistent in many cases. This consistency was 

especially apparent in the occupational medical exposure sections.

4.5 REGULATORY COMPLIANCE 

The Linde Site Profile attempts to use personnel monitoring data as much as possible, including 

film badge beta and gamma dose data and worker urinalysis results, to determine worker external 

and internal doses. The purpose of this attempt is to satisfy the requirements outlined in 

42 CFR Part 82, which specifies the hierarchy of data that are to be used in dose reconstruction. 

However, NIOSH has not always complied with the hierarchy of data required in §82.2 and its 

implementation guides for monitored workers. The following two examples show NIOSH’s 

attempt to use pertinent hierarchy of data, such as bioassay and film badge doses, for dose 

reconstruction. 

(1) For missed occupational external dose, the site profile uses a complex surrogate data 

system, involving film badge beta and gamma measurements, survey measurements, 

uranium solid sample results, and uranium source data, to represent potential missed 

doses to workers in different periods of Linde operation. However, the film badge beta 

and gamma measurements are limited in scope and quantity, and may not be reliably 

extendable to represent the entire period. 

(2) For missed occupational internal dose, the site profile does not use limited worker 

urinalysis data available to estimate potential missed doses to workers in different periods 

of operation. It uses mainly airborne uranium dust concentration data for estimating 

worker inhalation intakes and ingestion intakes. However, NIOSH did use the available 

urinalysis data as predictive data (Table 10 of the site profile) to evaluate the 

reasonableness of the estimated inhalation and ingestion intake values derived from the 

airborne uranium dust concentration data. 
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5.0 VERTICAL ISSUES 

SC&A has developed a list of issues regarding the Linde Site Profile, which relate to the five 

objectives defined in the SC&A site profile review procedure (SC&A 2004a). Some issues 

pertain to a particular objective, while others pertain to several objectives. A matrix relating the 

objectives and the relative importance of each issue is provided in Section 6.0 (also in Table 1-1). 

The issues identified as findings map into the four broad categories discussed in Section 5.1, and 

the issues identified as observations map into the two broad categories discussed in Section 5.2. 

Many of the issues raised are applicable to other DOE and AWE sites, and should be considered 

in the preparation and revision of their site profiles. 

5.1 DISCUSSION OF ISSUES 

5.1.1 Unsupported Assumptions and Significant Uncertainties in Information Used in this 

Site Profile 

SC&A has identified numerous assumptions or values used in missed dose estimates (both 

internal and external) in the site profile that are not adequately supported by explanation, data, 

technical studies, or references, making them appear somewhat arbitrary and without adequate 

technical bases. In some cases, an assumption is made or a value selected from a range of 

estimated values in order to bound a dose parameter, that is not entirely justified or explained. In 

other cases, an assumption or value selected is not seen by SC&A as bounding. In addition, 

NIOSH seems to have missed several buildings and areas that could have contributed to worker 

exposures. These are flaws that affect the credibility and validity of the missed dose estimates. 

The Linde Site Profile strongly conveys the impression that there are many significant unknowns 

and uncertainties in historical and operational information regarding the site, which was in 

operation since the Manhattan Project, with respect to its facilities, operations, service timelines, 

remedial activities, and radiation protection practices. For example, the site profile identifies 

only seven buildings that were used for the MED uranium processing project, while there is 

evidence of the existence of several others, where workers may have been exposed to radiation. 

Many former Linde workers at SC&A’s site expert interviews in Buffalo, New York, identified 

more buildings (Buildings 1, 2, 19, 39, 52, 54, 57, and 59) that had been involved directly or 

indirectly with the MED uranium processing operations. In addition, several exterior areas and 

spaces adjacent to Buildings 30, 75, and 76 were purportedly used for staging uranium ore or 

processed waste piles during and after the MED operations, and were not considered by NIOSH 

in evaluating potential exposures to workers. Furthermore, two buildings have uncertain status; 

it is not possible to definitively verify through known reference documents the existence of 

Buildings A and B during the MED operation period and, if they were there during that period, 

to identify their MED-related functions. 

Examples follow of NIOSH assumptions used to establish the historical framework for 

estimating missed occupational doses. Many qualifying statements and expressions of 

uncertainty appear, and these assumptions, which are not accompanied by adequate explanations 
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of their rationales, do not appear to be claimant favorable. The site profile states the following in 

the beginning of Section 3.0, Estimation of Internal Exposure, 1942–1954: 

This document assumes that Tonawanda Laboratory’s primary internal exposure 

occurred from October 1, 1942 to December 31, 1946. ..... Tonawanda 

Laboratory AEC work is assumed to have stopped radiation work after December 

31, 1946, although some workers might have visited the Ceramics Plant buildings. 

The exact levels of contamination remaining at the Ceramics Plant and the nature 

of worker activities in areas of residual activity are unknown for the standby and 

rehabilitation periods. ..... Continued lower level exposures to uranium progeny 

and to radon are assumed, because some radioactive waste was disposed on site 

and because initial cleanup was not completed until the end of 1954; however, for 

the Ceramics Plant the uranium exposures would have dominated during the 

1947 to 1954 period. [Emphasis added.] (Davidson 2005, p. 26) 

Section 3.1, Estimation of Particulate Intakes, notes that “the pre-1947 operational period intakes 

are reserved. Therefore, the pre-1947 information is provided only as a description of what the 

likely upper bound exposures might have been, and is not currently planned for use in Linde 

dose reconstruction” (Davidson 2005, p. 27). The site profile section discusses the pre-1947 

period (making many assumptions and frequently using expressions indicating uncertainty), and 

states that “although short-term exposures might have exceeded 300 MAC, it is very 

unlikely. ... [Nonetheless,] the assumption of air concentrations at 300 MAC seems adequate to 

provide a quick estimate of exposure” (Davidson 2005, p. 28). The site profile does not explore 

adequately which workers, under what conditions, may have received exposures exceeding 

300 MAC, but, overall, this assumption appears conservative for the pre-1947 period (and, in any 

event, is not used in the dose reconstruction). 

The site profile then moves to the period after ore processing operations and assumes, without 

any supporting data, that exposures were substantially lower. It is interesting also to note that the 

document assumes that most workers’ exposures are even lower, based on later data from Linde 

and from other sites. While the data from other sites may be applicable to Linde, the site profile 

does not demonstrate this: 

After the ore processing, Linde began a standby period. It was initially and 

arbitrarily assumed that exposures decreased to 1 MAC during the standby 

period at the Ceramics Plant, and that exposures decreased to 0.1 MAC at the 

Tonawanda Laboratory after cleanup in 1946 until the end of cleanup at the 

Ceramics Plant in 1954. Based on reviews of later air concentrations at Linde, 

and reviews of air concentration data from other sites, it is believed that most 

workers’ exposures would have been much lower during these periods. [Emphasis 

added]. (Davidson 2005, p. 28) 

After the standby period, the site profile states that the bounding exposure to workers is 

(arbitrarily) assumed to be 33 MAC for the Ceramics Plant. The following quotation illustrates 

the acknowledged degree of uncertainty in this assumption: 
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Beginning November 1, 1947 at Linde Ceramics, workers were assumed to be 

exposed to 33 MAC and it was assumed this exposure continued through cleanup 

in 1954. Uranium progeny are not included in this later period, because only 

refined uranium was used and because the dose from intakes of contamination left 

from earlier work would have been insignificant compared to the dose to uranium 

during operations. (Davidson 2005, p. 28) [Emphasis added.] 

The site profile section continues by making an apparently “claimant-unfavorable,” or at least 

poorly justified, assumption about the length of the workweek, “To simplify calculations, it 

assumed that the workweek was 40 work hours long during all years, although it is likely that 

the workweek for many was in excess of 40 work hours especially during the earlier years. The 

assumed air concentrations are sufficiently large to account for any differences in actual hours 

exposed” (Davidson 2005, p. 28) [Emphasis added]. Hence, as illustrated by the preceding few 

examples, the credibility of the site profile and its claim to be claimant favorable appear to be 

compromised by the many unsupported or poorly justified assumptions that are made throughout. 

These issues, as well as others, are discussed in more detail in the following sections. 

5.1.2 Occupational Internal Dose 

The Linde Site profile discusses different periods in the plant’s history, with different operational 

scenarios and exposure pathways in each period. One division is presented in Table 5-1: 

Table 5-1: Time Periods 

Period Also Called Name 

October 1, 1942 – September 14, 1947 Pre-1947 Operation and Standby Period 

September 15, 1947 – June 30, 1949 1947–1949 Production Period 

July 1, 1949 – December 31, 1954 1949–1954 Cleanup Period 

January 1, 1955 – Present After 1954 Post-Cleanup Period 

The site profile estimates missed occupational internal exposures to workers at the Linde 

Ceramics Plant and Tonawanda Laboratory from 1942 to the present time via two separate 

pathways; inhalation and ingestion. For the inhalation pathway, the site profile estimates the 

particulate intakes by workers from a set of very limited measured airborne uranium dust 

concentration data (Davidson 2005, p. 30). These air concentration data consist of airborne dust 

radioactivity measured in general areas and breathing zones associated with various production 

locations and tasks, radon breath data used to determine the amount of radium inhaled in the 

body (Davidson 2005, p. 31), and uranium progeny in the airborne dust (Davidson 2005, p. 31). 

For the ingestion pathway, the site profile estimates the worker intakes from the air concentration 

data by multiplying the air concentration (worker inhalation intake quantities in dpm per cubic 

meter) by a factor of 0.2 (Davidson 2005, p. 35). The site profile also estimates separately the 

missed internal doses for workers, considering different operational histories and exposure 

scenarios. 
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For the uranium progeny intake, the site profile estimates the inhalation and ingestion intakes by 

apportioning the air concentration value among uranium and its progeny using the alpha ratios of 

uranium and progeny. However, the site profile only applies this approach for uranium progeny 

to the Tonawanda Laboratory, not to the Linde Ceramics Plant. It further assumes that the 

uranium progenies were not present in the Linde Ceramics Plant after 1947 (Davidson 2005, 

Table 11, p. 37). 

The site profile relies heavily on a single document, Health Hazards in NYOO Facilities 

Producing and Processing Uranium: A Status Report, April 1, 1949, to determine air 

concentrations of uranium and its progeny (AEC 1949a) (Davidson 2005, p. 27). Section 3.1 of 

the site profile outlines the methodology of the NYOO study: 

In 1949, the Medical Division of the NYOO published a report on the health 

hazards at seven facilities that produced and/or processed uranium for the AEC. 

These facilities included Mallinckrodt Chemical Works, Harshaw Chemical 

Company, Linde Air Products, Electro Metallurgical Company, and Vitro 

Manufacturing Company. The AEC used the information on work tasks with 

measured air concentrations in breathing zones, general areas and process areas 

to determine average air concentrations weighted by exposure times and summed 

these time-weighted air concentrations to determine daily time-weighted average 

air concentrations by job categories. (Davidson 2005, p. 27)

NIOSH derives from the NYOO report a single, primary air concentration value of 33 MAC for 

the Linde worker inhalation intake beginning in November 1, 1947, and continuing through the 

cleanup period in 1954.2

 NIOSH represents this value as a conservative, claimant-favorable, 

upper bound, claiming the unlikelihood of anyone receiving a greater exposure for any 

significant length of time. The site profile offers the following guidance: 

(1) Linde Ceramics Plant 

• Pre-1947: reserved by NIOSH (short-term may exceed 300 MAC) (Davidson 2005, 

p. 28)3

• Standby Period (August 1, 1946–September 14, 1946): reserved (1 MAC) 

(Davidson 2005, p. 28)3

• Rehabilitation Period (September 15, 1947–October 31, 1947): reserved (see 

Section 5.1.2.1)3

• November 1, 1947–December 31, 1954: 33 MAC 

• January 1, 1955–Present: 100 dpm/y U (Davidson 2005, p. 72) 

2

 As noted earlier in this section, the site profile derives ingestion intake from the inhalation intake data. 

3

 These time periods are covered by the SEC petition (i.e., October 1, 1942, to October 31, 1947). 

Nevertheless, the site profile review includes these time periods for completeness. It is important to emphasize that 

the time periods following the period covered by the petition are of particular interest to this report. 
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(2) Tonawanda Laboratory 

• Pre-1947: reserved 

• Cleanup Period (August 1, 1946–December 31, 1946): reserved 

• January 1, 1947–December 31, 1954: 0.1 MAC 

• January 1, 1955–Present: 100 dpm/y U (Davidson 2005, p. 72) 

The site profile occupational internal dosimetry section (Section 3, p. 26) summarizes NIOSH’s 

review of existing historical documents and data related to worker internal exposures at both the 

Linde Ceramics Plant and the Tonawanda Laboratory. Data include urinalysis records, airborne 

concentrations, and radon breath measurements. In addition, NIOSH’s contractor, ORAU, has 

compiled a list of historical documents and data files for Linde operations from early 1942 to the 

present (the “ORAU database”). ORAU interviewed two past Linde management staff members 

to obtain first-hand Linde operational information (Dupree 1983). ORAU also interviewed six 

Linde workers to understand more about the site operational environment. 

5.1.2.1 Special Exposure Cohort Petition 

On October 12, 2005, NIOSH issued an evaluation report on the Special Exposure Cohort (SEC) 

Petition 00044, covering a class of atomic weapons employees at the Linde Ceramics Plant from 

October 1, 1942, through October 31, 1947. The SEC period begins at the start of the Operation 

and Standby Period, and somewhat overlaps the next period, The Production Period, as defined 

in the table in Section 5.1.2 of this SC&A review. NIOSH concluded that dose reconstruction 

could not be completed because of a lack of sufficient biological or air monitoring, and process 

and radiological source information to estimate internal exposures to workers. 

In anticipation of this SEC evaluation conclusion, NIOSH decided to categorize all data and 

parameters pertinent to estimating occupational internal dose for workers for the period from 

October 1, 1942, to October 31, 1947, as “reserved” in the site profile. Nonetheless, the site 

profile contains information from the period covered by the SEC Petition in order to present a 

whole picture to the dose reconstructors. 

5.1.2.2 Use of Air Concentration Data 

Following the time period covered by the SEC petition (see Section 5.1.2.1), the site profile uses 

a “bounding” air concentration of 33 MAC as the benchmark value to estimate missed inhalation 

intake for workers at Linde Ceramics Plant and Tonawanda Laboratory. The site profile 

indicates the basis for this assumption: 

Up until the time of the 1949 AEC report, surveys by the NYOO indicated that out 

of 648 exposed workers at these plants, 9% were exposed to uranium air 

concentrations greater than 125 MAC (greater than 6250 mg/m3), 9% were 

exposed at 25–125 MAC (1250–6250 mg/m3

), and 82% were exposed to less than 

25 MAC (less than 1250 mg/m3

). Linde’s maximum time weighted exposure 
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during this period was 33 or 32 MAC (the data in the report’s text and its graph 

differed). (Davidson 2005, p. 27)

Airborne dust concentration surveys were performed in October and November 1948 at the 

Linde Site. None of the 65 employees involved in the survey were exposed to dust concentration 

levels above 33 MAC. Accordingly, a default chronic dust loading of 33 MAC for the time 

period after October 1947 may appear to be a reasonable upper bound. However, the exclusive 

use of airborne uranium dust concentration data (air concentration) at Linde, or, for that matter, 

at any uranium processing facility, for missed occupational internal dose estimation may not be 

claimant favorable. Several technical studies, including the 2003 Y-12 study, Practical Use of 

Personal Air Sampling (PAS) Data in the Internal Dosimetry Program at the Y-12 National 

Security Complex (Snapp 2003), and the Nuclear Regulatory Commission’s NUREG-1400, Air 

Sampling in the Workplace (Hickey 1993), demonstrate that relying on air concentration data 

could lead to underestimating the worker intakes and, consequently, the internal exposures. The 

Y-12 study shows as high as a factor of 10 underestimation from using intakes derived from air 

concentration data rather than the preferred bioassay data. In addition, the sensitivity of survey 

instruments, locations of the air sampling, and air flow studies of the buildings are not 

considered in the Linde Site Profile; these factors impact the accuracy of the air concentration 

data. It is also important to note that the air concentration data used are based on results of 

random grab-air samples in general areas and breathing zones, but not on results of continuous 

area samples in high-risk or high-dose areas. Therefore, SC&A believes that using air 

concentration data only in the Linde Site Profile can lead to significant uncertainties in worker 

inhalation intakes and the possible underestimation of missed internal doses. Furthermore, the 

use of time-averaged measurement data rather than, say, 95th percentile data, may fail to capture 

worker exposures in areas of high uranium dust concentrations. The time-averaging issue is 

discussed later in this section. 

Other issues with the maximum time-weighted average approach include uncertainties about the 

length of a workday and number of overtime hours, and the maximum air concentration value. 

These issues must be addressed when considering time-weighting. For example, the Linde Site 

Profile presents in different places various workweek lengths, including 40 hours, 48 hours, and 

54 hours. This inconsistency is discussed further in Section 5.1.2.8. 

Therefore, SC&A believes that using the 33 MAC value may not be justified, even though it has 

been called the maximum time-weighted exposure during this period. The value of 33 MAC 

may not be claimant favorable for some workers, since it does not address acute or abnormal 

intake scenarios. Hence, it is not demonstrably claimant favorable and defensible. In fact, the 

site profile does not address or identify any acute or abnormal intake scenarios at Linde at all; it 

is difficult to believe without strong evidence that none occurred. 

In summary, SC&A has identified the following list of potential flaws with using a default 

airborne uranium concentration of 33 MAC for the time period of September 15, 1947–

December 31, 1954: 

(1) The measured air concentrations at the seven AWE facilities of the NYOO study were 

made in breathing zones, general areas, and process areas. The facility average air 
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concentrations were used instead of area-specific or task-specific air concentrations. 

Therefore, these concentrations are not representative of a specific task or job category at 

Linde and do not take into consideration job categories where the potential for exposure 

may have been well above 33 MAC. 

(2) The average air concentrations were time-weighted by exposure times and summed to 

determine daily time-weighted average air concentrations by job category. The resulting 

air concentrations may not be representative of a specific period of operation at Linde. 

(3) The NYOO report indicates that 9% of the workers in these seven AWE plants were 

exposed to uranium dust concentrations greater than 125 MAC, i.e., about 4 times the 

assumed 33 MAC bounding concentration used at Linde. This “bounding” air 

concentration value is not bounding for at least 9% of the workers, and, therefore, is not 

necessarily claimant favorable. 

(4) This site profile claims that the value of 33 MAC is the “maximum” time-weighted 

average air concentration value for Linde. However, NIOSH indicates that a worker 

might receive a short-term (acute) exposure, exceeding 300 MAC (Davidson 2005, p. 28). 

(5) NIOSH did not perform any uncertainty estimation and analysis for the air concentration 

data used to derive the maximum time-weighted average air concentration values, as 

required by 42 CFR Part 81.11. SC&A believes that this is not in compliance with the 

requirements of the regulation. 

(6) Several technical studies, including the Y-12 study and NUREG-1400, show that using 

air concentration data could lead to underestimating the worker intakes and subsequently 

the internal exposures to the workers. The Y-12 study shows as much as a factor of 10 

difference between intakes derived from urinalysis (bioassay) data and intakes derived 

from air concentration data, with the latter underestimating the exposure. 

(7) The site profile does not provide any information or analysis on the sensitivity of survey 

instruments used to measure airborne uranium dust concentrations, locations of the air 

concentration measurements, and air flow studies of the buildings. This information is 

needed in order to determine the validity of the air concentration data and the 

uncertainties associated with them. 

(8) NIOSH assumes the air concentration value of 1 MAC for the Linde Ceramics Plant 

during the standby period, and 0.1 MAC after the cleanup period for the Tonawanda 

Plant. These two assumptions are based entirely on the derived value of 33 MAC for the 

operational period. SC&A believes that this approach may not be claimant favorable. 

If inhalation values are at issue, then so are the derived ingestion values, since the latter are 

determined by multiplying the former by a factor of 0.2. There also appears to be problems with 

this approach to deriving ingestion intakes because (1) there may be little relationship between 

the contamination on surfaces and that in the air (as seems to be the case for many uranium 
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processing facilities), and (2) using air concentration data for determining worker inhalation 

intakes may also lead to the underestimation of missed ingestion doses. This issue is discussed 

at greater length in Section 5.1.2.5. 

With all these unknowns and shortcomings, the use of measured airborne uranium dust 

concentration data alone for estimating missed occupational internal dose is questionable. The 

derived missed worker intake values presented in Tables 11 and 12 of the site profile are 

considered by SC&A as not necessarily claimant favorable.

5.1.2.3 Urinalysis Data 

The site profile indicates that there are some urinalysis data reported for the Linde workers 

(Davidson 2005, p. 29). These data, however, are not used to estimate worker internal doses 

(from uranium intake). Instead, NIOSH uses very limited air concentration data, as discussed in 

Section 5.1.2.2 of this report. In Section 3.8 (Davidson 2005, p. 35), the site profile uses the 

urinalysis data cursorily together with some urinalysis data from other AWE or DOE facilities as 

predictive references to validate the upper-bound chronic exposure quantity of workers to 

uranium dust and its progeny as 300 MAC. However, the site profile does not adequately 

explain why the potentially valuable urinalysis data were not used for dose reconstruction, as 

required by 42 CFR Parts 81.5 and 81.6. 

In addition, using air concentration data only and neglecting urinalysis data for estimating 

worker inhalation intakes is not fully compliant with the required “hierarchy of data” approach 

stipulated in 42 CFR 82.42, which gives greatest importance to bioassay data. There are 

seventeen sets of urinalysis data for over 100 uranium workers in the ORAU Database for the 

period between December 16, 1947 and January 30, 1950. The air concentration data used in the 

site profile do not cover all periods of the Linde operation and suffer from the deficiencies 

discussed in Section 5.1.2.2 of this report, and, therefore, are deemed inadequate (see Finding 2). 

In spite of these shortcomings, NIOSH still decided to use only these air concentration data for 

occupational internal dose reconstruction. The urinalysis results should have been used by 

NIOSH for estimating worker inhalation intakes by Linde workers during that covered period of 

time, instead of using measured air concentration data from other periods of time. 

5.1.2.4 Breathing Rate 

The inhalation pathway is the primary uranium intake mechanism for the workers engaged in 

uranium processing operations at Linde. The site profile selects a breathing rate of 1.2 m3/hr for 

worker intake (Davidson 2005, p. 35), which corresponds to the assumption that workers were 

primarily involved in light exercise during the course of the day. A single value for the breathing 

rate, however, may not be consistent with the working conditions in the facility during the early 

years of operation, and is inconsistent with practices in other NIOSH site profiles, such as 

Mallinckrodt, Bethlehem Steel, Y-12, INL, SRS, and Hanford. NIOSH should consider a higher 

breathing rate for Linde, representative of more strenuous exertion. For example, ore movers 

held heavy burlap bags close to their chests as they moved the ore between boxcars and 

storage/production areas. They moved rapidly and lifted and carried heavy loads, suggesting that 
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the assumed breathing rate of 1.2 m3

/hr is too low. A minimum rate of 1.4 to 1.7 m3/hr would be 

more representative and should be used instead. In addition, NIOSH has not considered oronasal breathing (characteristic of heavy labor), which produces greater deposition in the lung 

than nasal breathing alone. 

5.1.2.5 Ingestion Rate 

There are several potential ingestion pathways for Linde workers, including eating contaminated 

lunches, drinking contaminated water, touching the mouth with contaminated hands, and oronasal ingestion. The Linde Site Profile assumes an overall uranium ingestion rate of 0.2 (i.e., 

20%) of the inhalation intake values for workers; “The amount of activity ingested on a daily 

basis can be approximated by assuming it to be 0.2 times the activity per cubic meter of air” 

(Davidson 2005, p. 35).

Since the inhalation intake is estimated by using air concentration data, SC&A believes that the 

NIOSH approach could lead to the underestimation of ingestion intake and eventual missed 

ingestion doses for Linde workers. This air concentration issue is discussed at length in 

Section 5.1.2.2. In addition, for uranium processing plants, the ingestion rate of uranium may 

not be proportional to the airborne uranium concentration, because the dust loading on surfaces 

may be more directly related to spills and the direct deposition of large flakes of uranium that 

were never actually airborne. 

5.1.2.6 Radon Exposure and Concentration 

The Linde Site Profile treats radon exposures in Section 3.5: 

During Ceramics Plant pre-production and initial production (which involved 

only domestic ore processing), the only source of radon was African ore 

processing at Tonawanda Laboratory. The indoor and outdoor radon 

concentrations to which Ceramics Plant workers were exposed were assumed to 

equal the outdoor concentration resulting from Tonawanda Laboratory work. No 

direct measurement of this was available. An estimate was made based on the 

lowest indoor concentrations measured at the Ceramics Plant during African 

ore processing. These were viewed as indicating the upper limit to the outdoor 

concentration since outdoor air is drawn indoors for ventilation. Approximately 

20% of the measurements in the Ceramics Plant ore processing building yielded 

results of 10 pCi/L or less, with most of these results at or near 10. Therefore, 

10 pCi/L was taken as the estimated outdoor concentration. 

(Davidson 2005, p. 32) 

Despite the above paragraph, SC&A is not clear why NIOSH uses the lowest indoor 

concentrations during African ore processing as the upper limit to both indoor and outdoor radon 

concentrations. With the presence of radium in the process areas and the waste produced, the 

radon levels could be quite high. 
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5.1.2.7 Raffinate Trace Radionuclides 

The dose consequences of raffinate trace radionuclides have not been adequately addressed in 

the Linde Site Profile. Raffinate contains Ac-227 and Pa-231, which are in the U-235 decay 

chain, as well as Th-230, which is in the U-238 decay chain. Inhalation of even small quantities 

of some raffinates, such as filter cake (one of the waste products at Linde), could result in 

significant doses to the workers. The issue of potential airborne contamination of raffinates must 

be more carefully assessed. In addition, NIOSH lists Ac-227 and Pa-231 in Tables 5, 11, and 12 

of the site profile as radionuclides of concern for internal exposure to Linde workers (Davidson 

2005, p. 37), but NIOSH does not include them in Table 39 for worker exposure during the 

cleanup period. In Tables 11 and 12, inhalation intake values are not listed for Th-230, Ac-227, 

and Pa-231 at all for the period from 1947 to 1954. NIOSH should further evaluate the potential 

exposure pathways for internal exposure of raffinate trace radionuclides, and investigate the 

relative impact of trace radionuclide intakes to the total dose. 

5.1.2.8 Assigned Weekly/Annual Work Hours 

The number of work hours used in calculating occupational internal and external doses for 

workers is inconsistent for different periods of Linde operations and, therefore, not claimant 

favorable, since the maximum is not used in most places. The site profile represents in Table 4 

(Davidson 2005, p. 24), as well as in several other places, that workers routinely worked more 

than 40-hours per week, and, in some cases, as many as 54-hours per week (i.e., six 9-hour days) 

for 50 weeks per year (2,700 hours per year). But in many instances, NIOSH uses the standard 

40-hours per week assumption for missed dose estimations. This approach is not only 

inconsistent, but also not claimant favorable. 

The following is a list of different weekly work hours or annual work hours assigned in the 

Linde Site Profile for missed occupational internal dose estimation (Davidson 2005): 

• Airborne uranium dust concentrations: 40-hours per week (p. 28) 

• Radon concentrations: 2,040-hours per year (p. 32) 

• Uranium inhalation intakes values: 2,000-hours per year (p. 35) 

• Uranium ingestion intake values: 8-hours per day, 250 days per year (p. 35) 

• Residual year occupational annual internal exposure: 2,000-hours per year (p. 72) 

These different and sometimes not claimant-favorable work-hour assumptions would 

underestimate worker inhalation and ingestion intake and, in turn, missed occupational internal 

dose to workers at Linde. NIOSH should use a set of consistent and claimant-favorable work 

hours. 

5.1.2.9 Surrogate Air Concentration Data 

Using the GM of air concentration data of seven AWE facilities in New York from a 1949 

AEC/NYOO report (AEC 1949a) as surrogate data to develop Linde site-specific worker 

inhalation intakes for the entire period of Linde Operation from 1942 to 1954 appears over-
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t

reaching and may potentially underestimate the missed occupational internal dose to workers. 

This approach is inappropriate, because the surrogate data are very limited and not representative 

of the actual Linde operation conditions, since ventilation was poor or non-existent and adequate 

radiation protection practices had not yet been developed in the earlier years of the Linde site 

operation. In addition, the Linde Site Profile uses daily weighted-average, not the upper bound 

or the 95th percentile values, of the exposure levels measured for workers as surrogate 

concentrations for the dust exposure levels of workers. As a result, the estimated missed internal 

dose to workers is not claimant favorable. SC&A believes that these surrogate data do not 

provide the best estimates for the reconstruction of missed occupational internal dose to workers 

for the period from 1942 to 1947. NIOSH should re-evaluate this approach. 

5.1.2.10 Geometric Values 

The statistical analysis approach used in the Linde Site Profile is not bounding and, most 

importantly, not claimant favorable. In Table 6 of the Occupational Internal Dose Section 

(Davidson 2005, p. 33), the site profile lists the GM or the GSD values for measured radon 

concentrations during African ore processing. First, there are no supporting calculations or data 

to show how these geometrical quantities were calculated. Second, these geometrical values 

would not provide “maximized” default values or assumptions to produce claimant-favorable 

worker doses. Third, NIOSH does not provide comparison of these geometrical values with 

NIOSH-prescribed 95th percentile values. For example, NIOSH’s use of the GSD values is not 

claimant favorable for routine day-long exposures of production workers in Linde process 

buildings. In addition, this statistical approach may not address very high, short-term, episodic 

airborne concentrations; short-term intakes during incidents; and intakes during the performance 

of tasks with a potential for high transient airborne concentrations. Since actual data were not 

available, 95th percentile, not GM, should be used to bound the potential inhalation and ingestion 

intakes, and to assign claimant-favorable missed occupational internal dose to workers. 

5.1.2.11 Lack of Comprehensive Uncertain y Analysis 

Providing geometrical values (GM and GSD) for inhalation intakes and radon concentration 

levels does not represent compliance with required uncertainty analysis. There are no 

uncertainties or errors from various sources (measurement, laboratory, and instrument) estimated 

for different assumptions, parameters, and factors used in the estimation of missed occupational 

internal dose in the Linde Site Profile. 

 

An assessment of uncertainties, as required by OCAS-IG-001 and OCAS-IG-002, has not been 

adequately developed for air concentration and radon measurement data used instead of bioassay 

data to assign internal dose. As described in the site profile, “little information was available” 

during the periods of production and non-production. In fact, the site profile uses different 

words to indicate that the information gathered is “uncertain,” such as “probably,” “likely,” and 

“assumes.” It gives the strong impression that the available data gathered are inaccurate and 

uncertain. Therefore, NIOSH should develop a method to determine best-estimates and their 

uncertainties, as well as the 95th percentile value of time-weighted values of inhalation intake, 

radon intake, and ingestion intake for internal dose calculations. 
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There are many assumptions and surrogate data used, with different degrees of uncertainty, in the 

estimation of occupational internal dose to workers for the period from 1942 to 1954. The 

accuracy and representation of these assumptions and surrogate data are questionable. Therefore, 

without associated uncertainty factors identified and applied, the eventual internal dose 

estimations are not properly supported. NIOSH should revise the Linde Site Profile to include 

uncertainty analysis, as required by 42 CFR 81. 

5.1.3 Occupational External Dose 

The Linde Site Profile uses a very complex scheme, combining film badge data, uranium solid 

sample analysis results, and facility radiation field measurements, to estimate missed 

occupational external dose to workers from 1942 to the present time. These data are limited, 

however, and not Linde facility/building-specific. Furthermore, the site profile uses different 

sets of data to estimate worker beta and gamma doses separately. 

The site profile provides a list of external dose data for the Linde Ceramics Plant and Tonawanda 

Laboratory, including the following sets: 

(1) 1949 Building 30 survey 

• Beta/gamma at contact and at 3’ (after vacuum cleaning and flushing) 

• Beta/gamma at contact and at 3’ (before decontamination) 

• Beta/gamma at contact and at 3’ (after decontamination) 

(2) 1949 April 19 and 22 surveys 

• Not clear whether the surveys performed for the whole site or just Building 30 

• Beta/gamma at 1 cm and gamma at 1 m (before vacuum cleaning and flushing) 

• Averaged gamma dose rate of 0.18 mR/hr about 4 times the estimated median contact 

gamma before decontamination (0.0438 mR/h from Table 13) 

• Averaged gamma dose rate of 0.87 mrep/h about 1.3 times the estimated median beta 

before decontamination (0.675 mrem/h from Table 13) 

(3) 1976 Building 30 survey 

• Not clear whether the survey performed for the whole site or just Building 30 

• Building 30 claimed to be the most contaminated building on site (not supported with 

document or data) 

(4) Pre-production (1942–1943) 

• Transport of uranium ore to Building 14 

• Indoor and outdoor exposures by Linde workers to beta/gamma sources possible 

• Used estimated Building 14 beta and gamma exposure rates before vacuum cleaning 

and flushing (“before”) in Table 13 

• Assumed these estimated “before” beta and gamma rates upper limit of outdoor levels 
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• Assumed exposure time of 0.5 hr/d, 6 d/wk, 50 wk/y for all site workers outdoors (i.e., 

150 hrs per year) 

• Assumed beta dose to whole body only for Linde workers 

• Assumed no extremity doses (i.e., ignoring ore movers) 

(5) Production (1943–1946) 

• No film badge measurements available for this period of Linde operations 

• Eight solid samples collected on January 23, 1944, for measurements of beta radiation 

(Davidson 2005, Table 19) 

• Dose rates assumed at contact (R/8-hr) (Davidson 2005, p. 44) 

(6) 1944 Radiation Surveys 

• March 2 and 3 surveys conducted 

• Identified two 200,000-lb piles and one 300,000-lb pile of 10% ore in the receiving 

room at the south end of Building 30 

• Identified one 30’ x 30’ x 20’ (~2,000,000-lb) pile of 8% ore in an unspecified 

location 

NIOSH has attempted to assign “somewhat conservative” assumptions in some portions of the 

external dose assignments, but these assumptions may not be bounding. Thus, the overall 

external dose approach results in reconstructed doses that are not necessarily claimant favorable. 

The important external dose assignments include: 

(1) For Production Period (1943–1946), the following external dose rates are assigned for all 

Linde workers:4

• A gamma exposure rate of 5.35 R/y for the whole body 

• An outdoor gamma rate of 0.020 R/y for all workers 

(2) For Rehabilitation and Production Period (1947–1949), the following external dose rates 

are assigned:4

• A beta dose rate of 1.95 rem/y for medium job category 

• A beta dose rate of 1.00 rem/y for low job category 

• A hands and forearm dose rate of 5.85 rem/y for medium job category 

• A hands and forearm dose of rate of 1 rem/y for low job category 

• An outdoor beta dose rate of 0.10 rem/y for the whole body for medium job category 

• A gamma exposure rate of 1.61 R/y for the whole body for medium job category 

• A gamma exposure rate of 0.48 R/y for the whole body for low job category 

4

 Note that portions of these time periods are covered by the SEC petition, but are addressed here and in the 

site profile for completeness. 
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• An outdoor gamma exposure rate of 0.020 R/y for all workers 

This complex scheme of estimating missed beta and gamma doses to Linde workers is not easy 

to follow. In many cases, there are no clear or, sometimes, any explanations why a particular 

assumption is made. Even when a clear explanation is presented, in many cases the rationale for 

the assumption is either inconsistent or not technically sound. These weaknesses lead SC&A to 

conclude that NIOSH’s external dose assumptions may not be claimant favorable, especially for 

the more highly exposed workers. 

An example of an unsupported assumption is the use of the factor of 3 shown in Table 13 

(Davidson 2005, pp. 39–40) to estimate the beta and gamma dose rates in Building 30 before 

vacuum cleaning and flushing in 1949. The site profile compares the very limited “survey data” 

taken in April 1949, before the vacuum cleaning and flushing of Building 30 to the estimated 

median contact gamma level before decontamination (Davidson 2005, p. 40). The comparison 

results show that the typical gamma dose rates measured at 3 ft were about 4-times higher than 

the estimated median contact gamma dose rate before decontamination. It is not clear why 

NIOSH decided to use the factor of 3 instead of 4. The site profile states the following: 

The measurements just before decontamination in 1949–1950 were made after the 

building had undergone vacuum cleaning and flushing. Brief, semi-quantitative 

reports were available for two one day surveys taken in April 1949, before the 

vacuum cleaning and flushing (Blatz 1949; Wolf 1949). Typical levels of gamma 

radiation measured at 3’ on April 19 and at contact on April 22 were similar and 

averaged about 0.18 mR/h, about four times higher than the estimated median 

contact gamma level before decontamination. Typical levels of beta measured at 

contact on April 22 averaged about 0.87 mrep/h, about 1.3 times higher than the 

estimated median beta level before decontamination. The April results were the 

basis of the estimate in Table 13 that beta and gamma radiation levels before 

vacuum cleaning and flushing were three times higher than the values measured 

afterward. (Davidson 2005, p. 40) 

The site profile proceeds from this point to use values developed based on the factor of 3 in 

Table 13 as the primary technical basis for missed beta and gamma estimates for several periods 

of Linde operations and for all Linde buildings. This further propagates the uncertainties 

inherently imbedded in the factor of 3, and potentially underestimates the missed beta and 

gamma dose rates to workers. 

Rather than comparing the median or average values for the external dose rate measurements 

before versus after decontamination, a ratio of the before versus after decontamination exposure 

rates could have been based on the upper 95th percentile confidence level of the means. 

Alternatively, the highest ratio observed at any given location could also have been used. In this 

way, there would be a higher level of assurance that pre-decontamination dose rates that are 

derived based on observed post-decontamination dose rates are claimant favorable. 
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5.1.3.1 Beta Dose 

Section 4.1.3.1 of the site profile covers beta dose to workers during the production years, 1943–

1946:5 

It is striking that the 1943–1946 dose rates indicated in Table 19 are much higher 

than the 1947–1949 rates discussed in Section 4.1.4.1. For 1943–1946 Step I 

Process Operators, the beta dose rate is estimated as 263 rem/y. For 1947–1949, 

the job with the highest beta dose rate was that of a Step III Process Operator, 

and film badge data indicate that the beta dose rate was only about 2 rem/y ... 

Despite the high estimates, the 1943–1946 data do not appear unreasonable. The 

dose rates in Table 19 are typical of contact dose rates for uranium materials (see 

Table A-1). The worker exposure times were debated and reviewed within the 

MED. In the MED’s interpretation of the data, with allowances for the protective 

measures (e.g., gloves) the rates were within the limit in effect at the time, 3.0 R 

per six-day week (Ferry 1944b) or 150 R/y. Therefore, the estimates in Table 19 

are considered a valid basis for dose reconstruction. (Davidson 2005, p. 44) 

For the period of 1943–1946, NIOSH assumes the worker beta dose rates were three times higher 

than those of the period of 1947–1949. According to the site profile (Davidson 2005, p. 45), “the 

factor three increase accounts for potential exposure to radiation from waste products from 

unrefined uranium ore and for the possibility that procedures in 1943–1946 did not involve as 

much radiological protection.” The basis of this assumption is that “unrefined uranium materials 

release approximately 2.6 times as much electron energy per uranium decay as refined uranium 

materials.” 

NIOSH establishes three exposure classes and determines the beta dose rates in Table 21 as 

follows: 

• High exposure category: 

– 263 rem/yr (Table 19) x 0.84 = 221 rem/yr for hands and forearms 

– 88 rem/yr (Table 19) x 0.84 = 74 rem/yr for remainder of body 

• Medium exposure category 

– 17.6 rem/yr for hands and forearms 

– 5.85 rem/yr for remainder of body 

• Low exposure category 

– 3 rem/yr for hands and forearms 

– 3 rem/yr for remainder of body 

5

 This time period is covered by the SEC petition. However, it is important to keep in mind that skin 

cancer, which could be caused by external beta exposures, is not one of the presumptive cancers covered by the SEC 

rule. 
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For outdoor beta dose rate determination, NIOSH decided to assign a value of 0.10 rem/yr, 

which is derived from the beta dose rate at 3 ft before vacuum cleaning and flushing (i.e., 

0.676 mrem/hr × 0.5 hr/d × 6 d/wk × 50 wk/yr). This value only addresses those working 

indoors all day. Those working outdoors about 8-hours per day should be assigned an outdoor 

beta dose rate of 1.6 rem/yr. 

Table A-1 of the site profile shows the beta surface dose rates from uranium-containing materials 

in the range of 200 mrad/hr, i.e., 400 rad/yr or rem/yr (using 2,000 work hours per year). 

Table 30, however, shows the beta doses for the Step III Process in Building 30 of the Linde 

Ceramics Plant (1947–1949) in the range of 2 rem/yr, i.e., 1 mrem/hr (using 2,000 work hours 

per year). The beta dose rate estimated for Step I Process Operations in 1943–1946 was 

263 rem/yr, which is less than the typical contact dose rates for uranium materials in Table A-1 

by a factor of close to 2. In addition, the assumption made by NIOSH that Linde radiation 

protection programs could effectively limit Linde workers within the MED guidelines of 3.0 R 

per 6-day week, or 150 R/yr, is not very convincing, due to the fact that Linde did not have much 

of a radiation protection program. None of the site experts SC&A interviewed in Buffalo 

recalled that they had received any type of radiation protection training while working at the 

Linde Site. 

In the Linde Safety Rules and Practices Handbook, there is only a very small section addressing 

radiation matter, titled “Radiation Areas.” It states the following (pp. 22–23): 

All of our radiation work is conducted under strict procedures set up by the State 

and Federal Governments. Special protective measures are taken to insure that 

none of our radiation workers are exposed to excessive amounts of radiation. 

All radiation work areas are clearly identified with radiation signs and/or lights 

and barricades for your protection. Do not enter these areas unless you have 

been specifically authorized to do so. 

Even though the estimates in Table 19 are close to the typical beta dose rates for uranium 

materials, they are not valid for bounding the potential beta doses workers might have received at 

the Linde Ceramics Plant and Tonawanda Laboratory during 1947–1949. In addition, the site 

profile attempts to use the dose values in Table 19 as surrogates for the entire period of 1943–

1946 (Davidson 2005, p.45): 

In order to use the data in Table 19 to estimate time-averaged beta dose rates 

applicable to the entire 1943–1946 production period at Linde, the ratio of the 

average dose rates to those measured in Table 19 was estimated. Table 20 

documents the determination of the ratio. The grade range of the L-30 ore used 

at Linde was 8–12% (Aerospace Corporation 1981, Table B-1). To obtain the 

highest ratio, it was assumed that the measurements in Table 19 were made on 

the lowest L-30 ore grade, 8%. It was also assumed that beta dose was 

proportional to electron energy released per decay and the worker doses were 

proportional to the mass of ore processed. With these assumptions, it was 
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estimated that average doses would have been 0.84 of the doses predicted by 

using the data in Table 19. Therefore, the estimated dose rates in Table 19 were 

multiplied by 0.84 to obtain time-averaged dose rates for 1943–1946 production. 

The results are in Table 21, which summarizes the results of all beta dose rate 

estimates in this section and groups the job categories into three groups (high, 

medium, and low).

NIOSH believes that by determining the ratio of the average dose rates of different types of ores 

to those measured in Table 19 for 8% L-30 uranium ores, the time-averaged beta dose rates can 

be derived for workers at Linde during the period from 1943–1946. The highest ratio was 

determined in Table 20 to be 0.84, which was selected by NIOSH for the application. The Linde 

Site Profile does not provide details about how the beta dose rates are time-averaged. 

SC&A has identified two potential flaws in this dose rate assignment. First, the dose rates in 

Table 19 are not necessarily bounding, because an individual could work with higher-grade 

uranium ore for a very long period of time in a processing area. Even when the highest ratio is 

used, there could be underestimation of the beta dose rates. Second, the time-averaged beta dose 

rates would not provide bounding dose rates for high-risk jobs. Therefore, this method and its 

determined beta dose rates may not be claimant favorable for all workers. 

From the application of the beta dose rates presented in Table 19, NIOSH proceeds to determine 

beta dose rates for various job categories of workers at the Linde Ceramics Plant and Tonawanda 

Laboratory. These job categories are also grouped into three exposure potentials—high, medium, 

and low. However, the site profile does not provide support or explanation of how these beta 

dose rates are determined for each of these job categories in the three exposure groups. 

Additionally, NIOSH does not identify which estimated beta dose rates in Table 19 were used to 

determine the beta dose rates in Table 21. 

5.1.3.2 Gamma Dose 

NIOSH compiled the following external dose data for use in the site profile: 

• 1946–1947 (Standby Period and is included in the time period covered by the SEC) 

– NIOSH assumes all Linde workers are a guard or a general worker remaining in 

office, production buildings, and outdoors 

– Gamma measurements (Howland 1946) in 6 locations in Building 30 at 1” from 

surface; 4 measurements read 0 R/8h, and 2 measurements 0.005R/8h (0.625 mR/h) 

near the atypically contaminated and radioactive ore dumping grill; NIOSH took 

indoor gamma and beta levels from a 1949 characterization (Table 13), before 

vacuum cleaning and flushing were performed (Davidson 2005). 

– Outdoor gamma/beta is assumed by NIOSH to be same as indoor rates; gamma and 

beta dose rates in offices and buildings were found to be zero; NIOSH uses exposure 

times of 9 h/d, 6 d/wk, and 50 wk/y; assumed beta extremity dose rate equals the beta 

dose rate to the whole body 
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• 1947–1949 (Rehabilitation and Production Period) 

 – Weekly film badge measurements of beta and gamma exposure available 

 – 6000 records from January 7, 1948, to December 12, 1949 

 – Beta dose unit is rem equivalent to shallow dose at 0.07 mm Hp 

 – Limit of detection was 35 mrem 

 – Outdoor beta dose was assumed to be 0.1 rem/yr 

 – Average beta level in Building 30 from Table 13 was 0.676 mrem/hr 

• Dose rate of 34.5 mrem/wk or 1.7 rem/y (assuming 8.5 h/d, 6 d/wk, 50 wk/y) 

• Equivalent to film badge LOD of 35 mrem/wk 

For 1947–1949, NIOSH assumes that all Linde workers, including office workers, would receive 

a gamma dose of at least 1 rem/y. In addition, NIOSH uses the following assumptions for dose 

reconstructions: 

• Medium job category: 1.95 rem/y (Davidson 2005, Table 30, highest category) 

• Low job category: 1 rem/y (Davidson 2005, Table 30, lowest badged worker) 

For extremity dose, NIOSH assumes the following: 

• Medium job category: 5.85 rem/y (i.e., 3 × 1.95 rem/y) 

• Low job category: 1 rem/y 

For gamma dose, NIOSH determines the following: 

• Limit of detection: 45 mR 

• Outdoor gamma dose: 0.02 R/y 

The average gamma level in Building 30 from Table 13 (Davidson 2005) in this period is 

0.131 mR/h, i.e., 6.7 mR/wk or 0.34 R/y (assuming 8.5 h/d, 6 d/wk, 50 wk/y). This gamma dose 

rate is less than the film badge LOD value of 45 mR/wk. Therefore, NIOSH assumes the 

following: 

• Medium job category: 1.61 R/y for the whole body (Davidson 2005, Table 32, highest 

from film badge) 

• Low job category: 0.48 R/y for the whole body (i.e., 0.34 + 0.14 from Davidson 2005, 

Table 32) 

The beta and gamma doses presented in Table 33 (Davidson 2005) for the Tonawanda 

Laboratory in the period from 1942–1946 are not supported. The Linde Site Profile does not 

provide a clear explanation of how these dose values are derived. 

 

For the period from 1949–1954 (Cleanup Period), NIOSH uses dose rates from Table 13 

(Davidson 2005), some film badge measurements of beta and gamma emitters during the 1948 

removal of equipment from Building 30, and some floor and wall radiation levels measured at 
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the start and end of the 1949/1950 decontamination of Building 30. The following is the NIOSH 

approach to estimating missed beta and gamma doses: 

• The film badge measurements of beta and gamma radiation in Table 17 (Davidson 2005) 

include floor and wall radiation, in addition to radiation from contaminated equipment. 

• Missed beta and gamma dose for cleanup workers should include also outdoor beta and 

gamma doses (Table 17), which equals estimated indoor floor and wall radiation levels at 

3 feet before vacuum cleaning and flushing presented in Table 13 (Davidson 2005). 

NIOSH assumes the following: 

• Cleanup support workers would receive a beta and gamma dose that equals half of the 

beta, gamma, and outdoor doses in Table 17 (Davidson 2005). 

• Non-cleanup workers would receive a beta and gamma dose that equals 5% of the gamma 

and outdoor gamma doses in Table 17 (Davidson 2005). 

• Cleanup workers would receive an extremity dose that is 3 times beta dose to the 

remainder of the body. 

It does not appear that NIOSH provides documentation supporting many of these assumptions, 

and it is not clear whether these assumptions are claimant favorable. 

For the period after 1954 (Post-Cleanup Period), NIOSH uses mainly 1976 Building 30 survey 

data. It is not clear how the beta and gamma annual dose rates are estimated in Table 39 

(Davidson 2005). 

Documents in the ORAU database contain gamma radiation rates measured in June 11, 1945, at 

the Linde Site from high grade ore stored in 55-gal steel drums (15-gauge, 1000 lbs). These dose 

rates, however, were not used in this Linde Site Profile. Sources of the data used by NIOSH in 

estimating the missed external (beta and gamma) doses are summarized as follows: 

• Eight solid samples of uranium material were collected in January 23, 1944, from Step I 

ore, ore tailings, barium cake, soda salt, and Step II iron cake. These samples were sent 

to laboratories of the Medical Section of the MED for radiological analyses. Results are 

compiled in Table 19 and used for estimating beta doses during the production period 

between 1943–1946. (Davidson 2005, p. 43) 

• Gamma film badge data were available from January 31, 1944, to February 26, 1945, 

except for a 3-week gap (April 18, 1944–May 8, 1944). Results are compiled in Table 23 

for five job categories (ball mill operator, loader, ore sampler, process operator, 

weighmaster) of the Step I African ore processing (January 31, 1944–November 12, 

1944) ,and in Table 25 for six job categories (ball mill operator, cleanup, loader, ore 

sampler, process operator, weightmaster) of the Step I Domestic ore processing 
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(November 13, 1944–February 26, 1945). These results are used for estimating missed 

gamma doses during the production period between 1943–1945 (Davidson 2005, p. 47). 

• Gamma radiation levels were measured at six locations inside Building 30 on October 22, 

1946. These measurements were made at 1 inch above the surface of interest. Results 

were reported as R/8 hours and are compiled in Table 16 for estimating beta and gamma 

doses during the standby period of the Linde Site (Davidson 2005, p. 41). 

• Weekly beta and gamma film badge data were available between 1947 and 1949 during 

the rehabilitation and production period at Linde Site. These weekly film badge data are 

categorized into different job categories in Tables 28 and 29. Results are compiled in 

Table 29 and Table 30 for estimating Step III worker beta doses, and in Table 31 and 

Table 32 for estimating Step III worker gamma doses during that period of operation 

(Davidson 2005, p. 54). 

• Survey measurement data from two 1-day surveys in Building 30 on April 19 and April 

22, 1949, were made before the building had undergone vacuum cleaning and flushing 

(Blatz 1949; Wolf, 1949). Results are compiled in Table 13 as the primary basis for 

estimating worker beta and gamma doses. The Linde Site Profile states on page 40, “The 

April results were the basis of the estimate in Table 13 that beta and gamma radiation 

levels before vacuum cleaning and flushing were three times higher than the values 

measured afterward” (Davidson 2005, p. 39). 

• Beta and gamma film badge data were available during the 1948 removal of equipment 

from Building 30. These data were determined by NIOSH to include floor and wall 

radiation, in addition to radiation from contaminated equipment. Results are compiled in 

Table 17 and Table 18 for estimating cleanup and non-cleanup worker beta and gamma 

doses for the cleanup period between 1949 and 1954 (Davidson 2005, p. 41). 

• Beta plus gamma survey measurement data from pre-decontamination (but after vacuum 

cleaning and flushing) and from post-decontamination of Building 30 in 1950 were 

available (Heatherton 1950). These measurement data were analyzed with the computer 

program LOGNORM4 to estimate the median measurements at 3 feet. Their median 

results are compiled in Table 13 for estimating worker beta and gamma doses (Davidson 

2005, p. 39). 

• Actual worker film badge data are available for the period from January 31 to November 

12, 1944. However, many of these film badge data were average exposures for the entire 

period. Therefore, they are not representative of a particular job condition, a particular 

operation, a specific job location, or high-dose jobs. SC&A does not consider this 

averaging approach to be claimant favorable, because the average exposures do not 

represent the upper-bound dose rates, especially since historical and operational 

information are scarce for the Linde Site. 
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NIOSH determines the highest worker gamma dose rate to be 5.35 R/yr for the Loader job 

category in Table 24 (Davidson 2005, p. 50), but this gamma dose rate is based on the median of 

a lognormal distribution and, therefore, not representative of the upper-bound dose rates. Even 

though NIOSH applies this “maximum” gamma dose rate for the whole body to all Linde Plant 

employees during the period from 1943 to 1946, it may not be claimant favorable for workers 

who potentially might have been exposed to higher gamma dose rates in certain job conditions. 

However, SC&A believes that this value is claimant favorable for all other Linde employees. It 

is also instructive to note that gamma dose rates presented in Table 24 are averaged over a 1-year 

period. This makes them even less representative of the actual exposure conditions for workers 

who might have worked partially in Linde buildings and operations. NIOSH may want to 

consider using the upper 95th percentile of the mean for each of the three worker categories as a 

more claimant-favorable strategy. Alternatively, if it is known that specific locations/job 

categories had consistently high-exposure potential, the upper 95th percentile value of the full set 

of measurements may be appropriate as a default value for those workers whose job 

category/location is not known. 

The outdoor gamma exposure rate of 0.02 R/yr (Davidson 2005, Table 13, p. 39) is assigned by 

NIOSH for all Linde workers. This rate is based on the exposure rate of 0.131 mR/hr and an 

exposure time of 0.5 hr/d for workers working indoors most of the working day. For those 

workers who worked outdoors the whole time during the day, the exposure rate should be 

0.32 R/yr. 

5.1.3.3 Neutron dose 

There were no neutron exposure monitoring or measurements performed for workers at the 

Linde Site, based on information provided in the site profile and documents in the ORAU 

database. Even though there were large amounts of uranium present in various processes of the 

Linde operation, the only potentially significant source of neutron exposure to workers would 

have come from the (α,n) alpha-neutron reaction in materials where uranium was mixed with 

elements of low atomic number, such as fluorine and oxygen. The site profile states the 

following on page 62 (Davidson 2005): 

Ceramics Plant (Buildings 30 and 38) personnel are assumed to be exposed to 

1/10 of the daily production amount of U3O8 or UF4 at a distance of 1 foot. This 

quantity of material is a claimant-favorable estimate of the time-averaged amount 

of material likely to have been close to the maximally exposed worker during a 

work shift. The factor of 10 reduction takes into account several factors: that the 

plant operated around the clock so that each shift dealt with only 1/3 of a day’s 

throughput; that many workers were involved in each type of operation so that 

each worked closely with only a portion of a shift’s throughput; that an individual 

working in the vicinity of a large quantity of material (e.g., barrels of finished 

product) would on the average have been much more than 1 foot distant because 

of the large volume it would have occupied; and that even a worker in a job that 

involved being close to large quantities of material also had other activities at 

larger distances from the source term.
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The calculated neutron dose rates and annual doses from the (α,n) sources are tabulated in 

Table 35 (Davidson 2005, p. 63). These dose rates and annual doses are listed separately for 

Building 30, Building 38, and Building 14. From these dose rates and annual doses presented in 

Table 35, NIOSH determines the annual neutron doses (rem) for various categories of workers 

(all workers, high, medium, low, cleanup, non-cleanup, research, office) during various periods 

of operation in Table 36. 

However, the calculation of these neutron dose rates and annual doses cannot readily be 

reproduced, due to lack of supporting data and information. In addition, annual doses used in 

Table 36 do not match the dose rates compiled in Table 35. NIOSH needs to provide more 

supporting information, so that the dose reconstructor can use the information presented. 

Table 36 summarizes all neutron-related data and is easy to use. However, it is pertinent to 

question whether these neutron data are claimant favorable and technically sound. SC&A did 

not find any information or discussion in the site profile regarding type and sensitivity of film 

badge and survey instruments used, measurement geometry, angular dependence, uncertainty, 

and correction factors. 

5.1.3.4 Film Badge Data 

The lack of complete beta and gamma film badge data for the period from 1942 to 1954 

represents that there is a great potential for unaccounted beta and gamma doses. Available film 

badge and other radiation survey data are summarized in Table 5-2, reproduced below. 

Table 5-2: Available Film Badge/Survey Data and Surrogated Data Used 

Year Beta Film 

Badge Data Gamma Film Badge Data Beta/Gamma Film Badge Data Surrogate Data 

1942 

1943 

1944 Step I process workers 

(January 31, 1944–

November 12, 1994) for 

African ore 

8 solid samples from 

various locations 

1945 Step I process workers 

(November 13, 1944–

February 26, 1945) for 

domestic ore 

1946 Survey measurements in 

six locations inside 

Building 30 

1947 Weekly measurements 

1948 Building 30 equipment 

removal; weekly 

measurements (6000 

records) 

1949 Weekly measurements 

(6000 records) 

*Building 30 survey 

measurements 
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Table 5-2: Available Film Badge/Survey Data and Surrogated Data Used 

Year Beta Film 

Badge Data Gamma Film Badge Data Beta/Gamma Film Badge Data Surrogate Data 

1950 Beta plus gamma 

measurements in Building 

30 

1951 

1952 

1953 

1954 

Post1954 Building 30 floor and wall 

radiation level 

measurements in 1976 

*Primary surrogate data source

NIOSH developed a complex scheme, including using pre-cleanup survey data for the preproduction period from 1942 to 1943, uranium solid sample data for the period from 1943 to 

1946, a 1-day survey data in six locations in Building 30 for the period from 1946 to 1947, two 

1-day pre-cleanup survey data after vacuuming and flushing in Building 30 for 1949, and postdecontamination survey data for 1950. SC&A believes the missed occupational external doses 

estimated by using this scheme may not be claimant favorable for all workers. This complex 

surrogate external dose rate scheme is summarized in Table 5-3 below. 

Table 5-3: Complex Surrogate External Dose Scheme*

Missed Beta Dose Missed Gamma Dose Missed Neutron 

Dose 

Linde Ceramics Plant (Buildings 30, 31, 37, 38) 

1942 Table 15 Table 15 NA 

1943 Table 21 Table 21 Table 35 

1944 Table 21 Table 21 Table 35 

1945 Table 21 Table 21 Table 35 

1946 Table 21 Table 21 Table 35 

1947 NK NK NK 

1948 p. 57/58 p. 59 Table 35 

1949 pp. 57/59 & Table 18 p. 59 & Table 18 NK 

1950 Table 18 Table 18 Table 18 

1951 Table 18 Table 18 Table 18 

1952 Table 18 Table 18 Table 18 

1953 Table 18 Table 18 Table 18 

1954 Table 18 Table 18 Table 18 

Post-1954 Table 18 Table 18 Table 18 

Tonawanda Laboratory (Building 14) 

1942 Table 33 Table 33 Table 35 

1943 Table 33 Table 33 Table 35 

1944 Table 33 Table 33 Table 35 

1945 Table 33 Table 33 Table 35 

1946 Table 33 Table 33 Table 35 

1947–1954 Table 33 Table 33 Table 33 

*All assigned external doses summarized in Table 36 (Davidson 2005, p. 64) 

NK-Not Known
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In addition, the use of the 1948 weekly film badge data for assigning both beta and gamma doses 

during the removal of equipment in Building 30 is not appropriate for the entire period from 

1949 to 1954. These beta and gamma dose assignments in Table 36 contain median weekly 

photon doses and weekly median electron doses for use for unmonitored workers from 1942 to 

1954. These assignments are not likely to capture the full range of external exposures during 

that time period. Table 36 does not appear to be representative of various facilities and job 

functions that defined Linde operations/processes. 

Another problem occurring in Table 36 (Davidson 2005, p. 64) is that some of the beta and 

gamma doses cannot be reproduced or traced back to their sources. For example, there is no 

explanation or discussion on how the 1947 and 1949 (beta/gamma/neutron) doses are calculated, 

since they are all based on 1947–1949 weekly film badge data presented in Table 29 and 

Table 31 of the Linde Site Profile. Hence, it is not apparent that the assigned missed beta and 

gamma doses presented in Table 36 are technically sound and claimant favorable. 

5.1.3.5 Survey Measurement Data 

Several sets of survey measurement data were used in the Linde Site Profile to calculate the 

missed beta and gamma doses for workers from 1942 to 1954. These survey measurements do 

not cover the entire period of Linde operation. SC&A believes that NIOSH should improve the 

use of these data, because significant gaps exist for time periods when workers were not 

monitored for external or internal exposure. In addition, NIOSH did not evaluate or attempt to 

evaluate the adequacy, uncertainty, and accuracy of these data, which further weakens the 

assigned missed worker beta and gamma doses for the Linde workers. 

There is no discussion or information regarding survey instrument geometry and sensitivity. 

NIOSH should provide angular dependence (anatomic geometry) correction factors for external 

gamma doses, particularly for low-photon energies, where the angular dependence of the 

sensitivity of the survey meter or dosimeter is most pronounced. These correction factors are 

used to account for, for example, the bias introduced by a dosimeter worn at the neck level and 

the higher doses received by tissues/organs below the waist. 

5.1.3.6 Time-Weighted Averages 

Similar to internal dose calculations (see Section 5.1.2.2 for discussion), time-weighted averages 

of external exposure values contain significant uncertainties and frequently fail to capture dose to 

workers in areas of high beta or gamma radiation fields. In the external dosimetry section of the 

site profile, NIOSH determines the time-weighted average beta and gamma radiation dose rates 

during the standby period from 1946 to 1947 by time-weighting the dose rates with average 

worker exposure times, and summing to determine annual time-weighted average by job 

category (Davidson 2005, p. 41). This approach could underestimate the dose rates for highdose or high-risk tasks at the Linde Site. In addition, the site profile estimates time-averaged 

beta doses for the production period from 1943 to 1946: 
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With these assumptions, it was estimated that average doses would have been 

0.84 of the doses predicted by using the data in Table 19. Therefore, the 

estimated dose rates in Table 19 were multiplied by 0.84 to obtain time-averaged 

dose rates for 1943–1946 production. The results are in Table 21, which 

summarizes the results of all beta dose rate estimates in this section and groups 

the job categories into three groups (high, medium, and low). 

(Davidson 2005, p. 45)

Even when the site profile claims that the “maximum” time-weighted average dose values are 

used to provide upper-bound dose values, they would not represent maximized dose values for a 

worker, and may have limitations when used for denial of claims; nor do they give claimants the 

benefit of the doubt in the face of uncertainties associated with high-risk jobs. Individual doses 

could be far greater than these averages, even when the job description and work locations are 

known. Procedures for estimating 95th percentile values, for instance, would need to be 

developed, in which the claimant is given the benefit of the doubt. The Linde Site Profile lacks 

the needed procedures for dealing with these uncertainties. 

Other issues with the maximum time-weighted average approach include uncertainties about the 

length of a workday, overtime hours, and the maximum air concentration value. These issues 

must be addressed when considering time-weighting. For example, the site profile presents 

various workweek lengths, including 40 hours, 48 hours, and 54 hours. This document uses 

inconsistent workweek lengths for different time-weighted averaging (Davidson 2005, p. 41). 

5.1.3.7 Contaminated Burlap Bags 

Olevitch (1944) reports the outdoor storage of contaminated ore bags that at 

times numbered in the thousands. In 1948, 1 mR/h gamma and 3,000 a dpm/ 

100 cm2 were measured from the soil in an area formerly used for storage of 

radioactive materials (Heatherton 1948h). (Davidson 2005, p. 22) 

During the SC&A-conducted interview in Buffalo, Linde site experts and past workers indicated 

that there were many thousands of used burlap bags stacked up in the open bay area behind 

Building 30 (see Attachment 3 of this review report). These bags were used for transporting 

uranium ore to Linde for processing. After the end of the operation period, the contaminated 

burlap bags were stored behind Building 30 awaiting disposal. Many Linde workers, operation 

staff, and administrative personnel sat on these contaminated bags during breaks and lunch 

periods. This practice continued for many years, exposing many people at close distances to beta 

and gamma radiation sources left over in the uranium-contaminated burlap bags. The site profile 

does not estimate the missed beta and gamma doses to workers resulting from sitting on or 

standing next to those burlap bags. 

5.1.3.8 Surrogate External Exposure Data 

The lack of complete film badge data from 1942 to 1954 at the Linde Site represents a period for 

which the potential for unaccounted beta and gamma doses is greatest. NIOSH’s use of pre-
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cleanup survey data for the pre-production period from 1942 to 1943, 8 solid ore samples data 

for the period from 1943 to 1946, data from a 1-day survey of 6 locations in Building 30 for the 

period from 1946 to 1947, data from two 1-day pre-cleanup surveys after vacuuming and 

flushing in Building 30 for 1949, and post-decontamination survey data for 1950 is complex, 

over-reaching, inadequately supported, and may not be claimant favorable for many workers. In 

addition, the use of the 1948 film badge data collected during the removal of equipment in 

Building 30 for assigning both beta and gamma doses for the period from 1949 to 1954 is not 

appropriate, because these data do not account for external doses to workers from contaminated 

burlap bags, contaminated soil, and other contaminated sources during the cleanup activities. 

The beta and gamma dose assignments presented in Table 36 contain median weekly photon 

doses and weekly median electron doses for unmonitored workers from 1942 to 1954. These 

dose assignments are not likely to capture the full range of external exposures to the workers. 

SC&A believes that assigned doses are not representative of potential exposures to workers in 

the many facilities, locations, and job functions that defined the Linde operations and processes. 

5.1.3.9 Assigned Weekly/Annual Work Hours 

In Table 4 and several other sections (Davidson, p. 24), the Linde Site Profile represents that 

personnel at Linde worked longer than the standard 40-hours workweek; as much as 9 hours per 

day, 6 days a week and 50 weeks per year. However, in calculating external exposure values, 

NIOSH inconsistently uses different work-hour values: 

• Pre-production beta and gamma radiation levels (6 days per week, 50 weeks per year, 

Table 15, p. 40) 

• Standby beta and gamma radiation levels (9 hours per day, 6 days per week, 50 weeks per 

year, Table 16, p. 41) 

• Cleanup beta and gamma radiation levels (8.5 hours per day, 5/6 days per week, 

50 weeks per year, Table 18, p. 43) 

• Production beta radiation levels (up to 50 hours per week, Table 19, p. 44) 

• Production year outdoor beta dose rates (6 days per week, 50 weeks per year, p. 46) 

• Production gamma radiation levels (48 hours per week, p. 48) 

• Production year outdoor gamma dose rates (6 days per week, 50 weeks per year, p. 54) 

• Tonawanda Laboratory beta and gamma radiation levels (8.5 hours per day, 6 days per 

week, 50 weeks per year, Table 33, p. 60) 

• Neutron exposure rates (8 hours per day, 6 days per week, 50 weeks per year, p. 61) 

SC&A believes that applying these different work hours to the missed occupational external dose 

estimation would underestimate the missed dose or exposure assignments. NIOSH should use a 

set of consistent and claimant-favorable work hours for use in the dose reconstruction. 
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5.1.3.10 Use of Geometric Mean Values 

Similar to the practice followed in the Occupational Internal Dose section of the site profile, the 

statistical approach used in the Occupational External Dose section may not be claimant 

favorable for many workers. In Tables 13, 14, 15, 17, 18, 21, 23, 24, 25, 26, 29, 30, 31, 32, and 

33, the site profile lists the GM or the GSD values for various assigned external exposure values. 

There are no supporting calculations or data to show how these geometrical quantities were 

calculated. In addition, the applicability of these distributions to all job categories is not 

immediately apparent. In general, NIOSH’s use of the GSD approach may not be claimant 

favorable for routine day-long exposures of some production workers. The NIOSH approach 

may not address very high, short-term, episodic exposures and short-term exposures during 

incidents. Since actual data were not available, the cited distributions may not be applicable to 

some workers. 

5.1.3.11 Lack of a Comprehensive Uncertainty Analysis 

As described in the site profile, “little information was available” during the periods of 

production and non-production. In fact, the site profile uses different words to indicate the 

information gathered is “uncertain,” such as “probably,” “likely,” and “assumes.” It gives the 

strong impression that the available data gathered are inaccurate and uncertain. Therefore, 

NIOSH should develop a more comprehensive characterization of the uncertainties in external 

exposures, as described in OCAS-IG-001. 

There are many assumptions and surrogate data used, with different degrees of uncertainty 

associated with them, in the estimation of occupational external dose to workers for the period 

from 1942 to 1954. The accuracy and representation of these assumptions and surrogate data are 

questionable. Therefore, without associated uncertainty factors identified and applied, the 

eventual external dose estimates are not adequately supported. NIOSH should revise the Linde 

Site Profile to include an uncertainty analysis, as required by 42 CFR 81. 

5.1.4 Occupational Environmental Dose 

The Linde Site Profile does not provide a separate section on potential missed occupational 

environmental dose to workers, and does not address or systematically discuss any potential 

worker environmental exposure pathways for the entire period from 1942 to the present time. 

However, in some of the site profile sections, NIOSH does identify some potential worker 

exposure to outdoor radon activities and gamma radiation levels in different periods of operation. 

During the site interview, Linde site experts and past workers indicated that there were 

significant environmental contaminations in external areas of Linde Site, including the following: 

• Contaminated pond and stream 

• Contaminated parking lots 

• Contaminated stormwater sumps 

• Contaminated underground tunnels 

• Contaminated rooftops 
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• Contaminated burlap bags 

• Ore tailing piles 

• Process waste piles 

• Incineration exhausts 

• Contaminated injection wells 

• Burial plots 

The site profile provides discussions of some of these sources of environmental contamination as 

follows: 

Olevitch (1944) reports the outdoor storage of contaminated ore bags that at 

times numbered in the thousands. In 1948, 1 mR/h gamma and 3,000 a dpm/ 

100 cm2

 were measured from the soil in an area formerly used for storage of 

radioactive materials (Heatherton 1948h). (Davidson 2005, p. 22) 

Notes from the Worker Outreach Meeting on April 18, 2005 mention 

contamination associated with Building 57, and an additional review of the BNI 

1993 remedial investigation report shows areas of residual radioactive 

contamination were associated with areas in or near Buildings 57, 58 and 90. 

The highest indoor radiation levels were found in the principal production 

buildings, 30 and 38. Linde was designated as a FUSRAP site in 1980. 

Additional radiological surveys and decontamination efforts followed (BNI 1993). 

These led eventually to demolition of Building 14 and all of the Linde Ceramics 

buildings involved in MED/AEC work except Building 31. Table 1 shows 

demolition dates. As of 2004, Building 31 remained in use, and onsite soil 

remediation was in progress with completion scheduled for 2007 (Pilon 2004). 

(Davidson 2005, p. 26) 

5.1.4.1 Outdoor Doses 

Although the Linde Site Profile does not address missed occupational environmental doses to 

workers, NIOSH does evaluate several potential outdoor beta and gamma sources. In some cases, 

however, NIOSH ignores the outdoor doses (Section 4.1.3.1.2, p. 46; Section 4.1.3.2.2, p. 54) 

after the doses are calculated. In other cases, NIOSH uses them in the external dose calculations 

(Table 32, OD added, p. 59). The site profile also evaluates outdoor radon concentrations, but 

not on direct measurement (Section 3.5.1, p. 32), and presents some outdoor airborne 

radioactivity data for July 2000 through June 2004 (Davidson 2005, Table 38, p. 72). 

Potentially, the missed occupational environmental doses to workers are small, as compared to 

missed occupational internal and external doses. As a principle of claimant-favorability and 

completeness of dose reconstruction required by EECIOPCA, however, NIOSH should evaluate 

the potential missed environmental doses to production workers, support workers, administrative 

personnel, cleanup workers, and other facility workers in different periods of the Linde operation 

from 1942 to the present time. 
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5.2 OTHER OBSERVATIONS 

In addition to the specific findings and observations that appear in other sections of this review 

report, there are several general observations that can be made. 

• The content and organization of the Linde Site Profile are quite uneven, where useful 

assumptions, default values, and specific guidance for dose reconstructors are often 

buried among general discussions about site activities, history, and technical information 

about dosimetry, which may or may not have practical application. 

• This site profile does not provide examples to assist dose reconstructors in using recorded 

records, missed dose assignment, and dose assignments when dosimeters read zero dose. 

• It is critical for NIOSH to conduct interviews with former workers and other site experts 

and integrate first-hand experience and/or association with Linde Site, so as to provide 

further insights on job category information, site practices, processes and conditions, 

management practices, and data integrity. First-hand experience and association with the 

Linde facilities enable site experts and workers to provide original perspectives and 

information concerning site practices and exposure histories that may not appear in the 

official records. NIOSH tends to conduct worker outreach meetings after the site profiles 

are completed. 

• The site profile does not provide much information on Linde worker records or files. 

During the site interviews, many site experts indicated that they had not seen their own 

dosimetry or personnel records. They believe that there appear to be many incident 

reports, occurrence reports, contamination reports, and worker uptake reports that were 

not included in the records. NIOSH should look into the possibility of many missing 

dose records in Linde worker files. The suspected incompleteness of the worker records 

is a serious issue, since it may lead to significant underestimation of workers’ radiation 

doses. 

• NIOSH should provide angular dependence (anatomic geometry) correction factors for 

external gamma doses, particularly for low-photon energies, where the angular 

dependence of the sensitivity of the dosimeter is most pronounced. These correction 

factors are used to account for, for example, the bias introduced by a dosimeter worn at 

the neck level, and the higher doses received by tissues/organs below the waist. 

• NIOSH did not evaluate soil-sampling data at different facilities to determine potential 

worker intakes from resuspension of radioactive materials deposited on facility grounds 

and fugitive emissions from radiologically contaminated soil piles. NIOSH should 

consider an exponential model for resuspension intakes that takes account of the gradual 

increase in ore dust levels or radon levels at Linde facilities.
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6.0 OVERALL ADEQUACY OF THE LINDE SITE PROFILE AS A 

BASIS FOR DOSE RECONSTRUCTION 

The SC&A site profile review procedure calls for both a “vertical” assessment (Section 5) of a 

site profile for purposes of evaluating specific issues of adequacy and completeness, and a 

“horizontal” assessment of how the profile satisfies its intended purpose and scope. This section 

addresses the latter objective by evaluating (1) how, and to what extent, the site profile satisfies 

each of the five objectives defined by the Advisory Board for ascertaining adequacy; (2) the 

usability of the site profile for its intended purpose (i.e., to provide a generalized technical 

resource for the dose reconstructor when individual dose records are unavailable); and 

(3) generic technical or policy issues that transcend any single site profile that need to be 

addressed by the Advisory Board and NIOSH. As mentioned in the Introduction, the practice of 

addressing the same items from several different perspectives has led to some redundancy in the 

report. 

6.1 SATISFYING THE FIVE OBJECTIVES 

The SC&A review procedure, as approved by the Advisory Board, requires that each site profile 

be evaluated against five measures of adequacy; (1) completeness of data sources, (2) technical 

accuracy, (3) adequacy of data, (4) site profile consistency, and (5) regulatory compliance. The 

SC&A review of the Linde Site Profile finds that several shortcomings and potential issues of 

varying significance need to be addressed. Many of the issues involve the use of unsupported or 

unexplained assumptions in the dose estimates, a lack of sufficient conservatism or consistency 

in some key assumptions or estimation approaches, incomplete analyses of data, incomplete 

reflection of operational or dosimetric history, or a lack of regulatory compliance with the 

hierarchy of data and uncertainty analysis requirements. Key issues are summarized below and 

in Table 6-1 (a duplicate of Table 1-1), which provides a matrix representation of the identified 

issues sorted according to the SC&A findings and observations. Detailed discussions of these 

issues are provided in Section 5 of this report. 

An “X” in the table indicates significant shortfalls in meeting the corresponding review 

objectives for the indicated topics in the site profile. These shortfalls have been discussed either 

within the text of the findings themselves, or, in many cases, in special sections that address one 

or more of these shortfalls. The first column of the table indicates the primary place within the 

report that treats each issue. The last column of the table presents three categories of potential 

related regulatory non-compliance concerns for the listed issues. These three categories are 

defined again briefly as follows: 

• Category 1: Least challenged by any deficiencies in available dose or monitoring data 

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

corresponds to a probability of causation (POC) value in excess of 50%, and assures 

compensability to the claimant. 

• Category 2: The use of upper-bound values is limited to those instances where the 

resultant maximized doses yield POC values below 50%, which are not compensated. 
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For this second category, the dose reconstructor needs only to ensure that all potential 

internal and external exposure pathways have been considered. 

• Category 3: The most complex and challenging dose reconstruction represents claims 

where the case cannot be dealt with under one of the previous two categories and a 

detailed analysis is required. 

Table 6-1: Issue Matrix for the Linde Site Profile 

Descriptions (a)

Issue 

Classification 

Objective 1: 

Completeness 

Of Data 

Objective 2: 

Technical 

Accuracy 

Objective 3: 

Adequacy of 

Data 

Objective 4: 

Site Profile 

Consistency 

Objective 5: 

Regulatory 

Compliance 

Issue 1: (5.1.1) Unsupported 

Assumptions and Significant 

Uncertainties in Information Used 

Finding (1) X X 

Issue 2: (5.1.2.2) Use of Air 

Concentration Data Finding (2) X X X X 

Issue 3: (5.1.2.3) Urinalysis Data Finding (3) X X X X 

Issue 4: (5.1.2.2) Time-Weighted 

Averages Finding (4) X X 

Issue 5: (5.1.2.4) Breathing Rate Observation X X 

Issue 6: (5.1.2.5) Ingestion Rate Observation X X 

Issue 7: (5.1.2.6) Radon Exposure and 

Concentration Observation X 

Issue 8: (5.1.2.7) Raffinate Trace 

Radionuclides Finding (5) X X 

Issue 9: (5.1.2.8) Assigned Work 

Hours Finding (6) X X 

Issue 10: (5.1.2.9) Surrogate Air 

Concentration Data Finding (7) X 

Issue 11: (5.1.2.10) Use of Geometric 

Mean Values Finding (11) X X 

Issue 12: (5.1.2.11) Lack of 

Comprehensive Uncertainty Analysis Finding (9) X X X 

Issue 13: (5.1.3) Complex Missed 

External Dose Surrogate System Finding (8) X X X 

Issue 14: (5.1.3.4) Film Badge Data Finding (8) X 

Issue 15: (5.1.3.5) Survey 

Measurement Data Finding (8) X 

Issue 16: (5.1.3.6) Time-Weighted 

Averages Finding (4) X X 

Issue 17: (5.1.3.7) Contaminated 

Burlap Bags Observation X 

Issue 18: (5.1.3.8) Surrogate External 

Exposure Data Finding (7) X 

Issue 19: (5.1.3.9) Assigned Work 

Hours Finding (6) X X 

Issue 20: (5.1.3.10) Use of Geometric 

Mean Values Finding (11) X X 

Issue 21: (5.1.3.11) Lack of 

Comprehensive Uncertainty Analysis Finding (9) X X X 

Issue 22: (5.1.4.1) Outdoor Doses Finding (10) X X 

(a) Report section numbers discussing the issues are given after the issue number. 

6.1.1 Objective 1: Completeness of Data Sources 

The breadth of data sources used as a basis for the Linde Site Profile is evident in the 153 

reports, papers, and other documents cited as references, including a number of authoritative 

historical documents dating back to the start of site operations in the 1940s. In addition, there 
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are other technical and regulatory documents gathered by NIOSH that are deposited in the 

ORAU database. Based on a review of the Linde Site Profile (Davidson 2005), it is evident that 

NIOSH compiled information and characterized activities and operations in seven buildings and 

processes at the Linde Site. In fact, this review cites, in Section 4.0 of this report, the 

compilation of worker job categories and high-risk jobs as strengths. Also noteworthy is the use 

of the film badge beta and gamma measurement data, in addition to uranium solid sample results 

and survey measurements, as part of the missed occupational external dose estimation process. 

Notwithstanding the general adequacy of the data sources, SC&A has identified a number of 

areas as deficient: 

(1) NIOSH did not adequately use the available worker urinalysis data as part of the missed 

occupational internal dose estimate. There are a total of 17 sets of urinalysis data in the 

collected Linde documents (Linde 2004a and 2004b) for the period from December 16, 

1947, to January 31, 1950. Even though NIOSH did use these urinalysis data for 

predictive comparison, the neglect of these data is deemed deficient. 

(2) NIOSH did not adequately compile historical data on medical x-ray equipment and 

techniques used at Linde Site. The Occupational Medical Dose section (Site Profile 

Section 5) is descriptive in the requirements, but lacking in substance. It provides no 

information on chest x-ray equipment used, radiographic film used, photoflurographic 

film used, pelvis/lumbar spine x-ray equipment used, or worker medical records. In the 

end, the site profile instructs dose reconstructors to use the NIOSH TIB ORAUT-OTIB0006, Dose Reconstruction from Occupationally Related Diagnostic X-Ray Projections, 

as the basis document for reconstructing occupational medical dose for workers. Overall, 

SC&A believes that using the default medical doses in ORAUT-OTIB-0006 is claimant 

favorable for the purpose of Linde dose reconstruction. 

(3) NIOSH did not develop missed occupational environmental dose to workers, even though 

the site profile provides some information pertinent to this exposure pathway. This 

deficiency is considered minor due to the relatively low potential exposures expected 

from this pathway. 

(4) Using the GM of 1949 air-concentration data as surrogate intake values for the entire 

period of 1942–1948, and using results of 8 solid samples from a 1944 report as surrogate 

beta dose values for the entire period of 1943–1946, may not be representative of, and 

may underestimate, the missed dose to some workers. This approach of applying 

narrowly defined exposure data as surrogate data to estimate a broad range of missed 

internal or external doses for Linde workers is inappropriate, because the applied data are 

very limited and not representative of the actual operation conditions. In addition, the 

assigned missed internal and external doses do not appear to be claimant favorable for 

workers who had a potential to experience elevated internal and external exposures. 

(5) SC&A found a lack of characterization of contaminated soil or process waste materials 

stored outdoors at Linde buildings, such as Building 30, in the site profile. For example, 
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large quantities of ore stored in four ore piles in early 1944 onsite at Linde were not 

treated. 

(6) SC&A found a lack of characterization of potential worker exposures at the waste tank 

farm, and in remediation and waste management in general. The list of radionuclides 

provided for those operations is incomplete and increases the potential for missed dose. 

(7) SC&A found an inadequate characterization of worker internal exposures. The site 

profile did not evaluate the worker urinalysis data for the period from December 16, 1947, 

to January 31, 1950. Even though these personnel monitoring data do not cover the more 

problematic early years of the Linde site operation from 1942 to 1947, the use of these 

data would not be supportive for the dose reconstruction. Instead, the Linde site profile 

uses very limited measured airborne uranium dust concentrations, and applies them to the 

entire period of operation. SC&A believes that these data are not sufficient to reconstruct 

missed internal doses. 

(8) The site profile does not characterize or provide any information on the potential missed 

worker external doses due to extremity exposure, skin contamination, and whole-body 

exposure to residual contamination in used burlap bags stored in different locations, such 

as Building 30. Page 22 of the site profile indicated thousands of contaminated burlap 

bags were stored in Building 30. 

(9) The Linde Site Profile lacks characterization of the potential missed internal and external 

doses for workers, who had performed maintenance work in the Linde underground 

tunnel system in the post-1954 period. These tunnels might have been contaminated 

cumulatively by storm-water runoff and melting snow water that washed down uranium 

dust and contaminated process waste piles during different periods of Linde operation. 

There is no indication that there were any radiation surveys or cleanup performed inside 

these tunnels. SC&A believes that this constitutes a deficiency in data collection. 

(10) SC&A found that the site profile is deficient in evaluating many potential exposure 

pathways to workers, including rooftops and parking areas contaminated by uranium dust 

emitted from unfiltered ventilation systems on top of Buildings 14, 30, 31, 37, and 38. 

6.1.2 Objective 2: Technical Accuracy 

There are a number of issues identified in the course of this review that may be classified as 

deficiencies in technical accuracy: 

(1) There are deficiencies in evaluating dose consequences of raffinate trace radionuclides in 

the Linde Site Profile. Raffinate contains Ac-227 and Pa-231, which are in the U-235 

decay chain, as well as Th-230. Possible doses from raffinate-related exposures have not 

been evaluated in the site profile. Inhalation of even small quantities of some raffinates, 

such as filter cake (one of the waste products at Linde), could result in significant doses 

to the workers. The issue of potential airborne contamination of raffinates must be more 
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carefully assessed. In addition, NIOSH lists Ac-227 and Pa-231 in Tables 5, 11, and 12 

of the site profile as radionuclides of concern for internal exposure to Linde workers 

(Davidson 2005, p. 37), but does not include them in Table 39 for worker exposure 

during the cleanup period; NIOSH should address these issues. In Tables 11 and 12, 

inhalation intake values are not listed for Th-230, Ac-227, and Pa-231 for the period from 

1947 to 1954. NIOSH should further evaluate the potential exposure pathways for 

internal exposure of raffinate trace radionuclides, and investigate the relative impact of 

trace radionuclide intakes to the total dose. 

(2) Using the GM of 1949 air-concentration data as surrogate intake values for the entire 

period of 1942–1948, and using results of 8 solid samples from a 1944 report as surrogate 

beta dose values for the entire period of 1943–1946, may not be representative of, and 

may underestimate, the dose to some workers. This approach of applying narrowly 

defined exposure data as surrogate data to estimate missed internal or external dose for 

Linde workers is inappropriate, because the applied data are very limited and may not be 

representative of the actual operating conditions. In addition, the assigned missed 

internal and external doses do not appear to be claimant favorable for many workers. 

(3) Time-weighted averages of internal and external exposure values contain significant 

uncertainties and frequently fail to capture dose to workers in areas of high uranium dust 

concentrations. Even using the maximum time-weighted average dose values would not 

represent maximized dose values and may have limitations when used for denial of 

claims; nor do they give claimants the benefit of the doubt in the face of uncertainties. 

Individual doses could be far greater than these averages, even when the job description, 

work location, and work duration are known. Procedures for estimating 95th percentile 

values, for instance, would need to be developed in which the claimant is given the 

benefit of the doubt in the face of significant uncertainties. The Linde Site Profile does 

not evaluate these potential uncertainties, and lacks the required procedures for dealing 

with them. Other issues include uncertainties about the length of a workday and the 

length of a workweek, as well as overtime hours; these must be addressed when 

considering time-weighting (see Finding 7). For example, NIOSH (Davidson 2005, p. 27) 

uses the maximum time-weighted exposure of 33 MAC instead of the maximum 

exposure of 125 MAC, even though it is noted that 9% of the exposed workers 

experienced the maximum exposure. This approach may not give many workers the 

benefit of the doubt (see Finding 1, also). 

(4) The number of work hours assumed in calculating occupational internal and external 

doses for workers is inconsistent for different periods of Linde operations and, therefore, 

may not be claimant favorable. The site profile has represented in Table 4 (Davidson 

2005, p. 24), and in many other places, that workers had longer workweeks than 40 hours, 

and, in some cases, worked as much as 9 hours per day for 6 days a week and 50 weeks 

per year. But, in most instances, NIOSH uses the standard 40 hours per week assumption 

for the missed dose estimation. This approach is inconsistent and not claimant favorable 

for some workers. 
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(5) The Linde Site Profile does not evaluate soil-sampling data at different facilities to 

determine potential worker intakes from resuspension of radioactive materials deposited 

on facility grounds and fugitive emissions from radiologically contaminated soil piles. 

NIOSH should consider an exponential model for resuspension intakes that takes account 

of the gradual increase in ore dust levels or radon levels over time.

(6) After the end of the operation period, many thousands of contaminated burlap bags were 

stored behind Building 30 awaiting disposal. Many Linde workers, operations staff, and 

administrative personnel sat on these bags during breaks and lunch periods over many 

years. These personnel had been exposed at close distances to beta and gamma radiation 

sources left over in those uranium-contaminated burlap bags. The site profile does not 

estimate the missed beta and gamma doses to workers resulting from sitting or standing 

next to the contaminated burlap bags.

(7) The Linde Site Profile assumes a breathing rate of 1.2 m3

/hour for worker intake. This 

value implies that workers were primarily involved in light activities during the course of 

the day. A single value, however, may not be consistent with the working conditions in 

the facility during the early years of operation, and is inconsistent with other NIOSH site 

profiles, such as for Mallinckrodt, Bethlehem Steel, Y-12, INL, SRS, and Hanford. 

NIOSH should consider a higher breathing rate of 1.7 m3/hour for the Linde dose 

reconstruction, which is more consistent with heavy labor from operations, such as 

hauling heavy burlap bags of ore. 

(8) The Linde Site Profile assumes an overall uranium ingestion rate of 0.2 of the inhalation 

intake values for workers. Since the inhalation intake is estimated by using air 

concentration data, SC&A believes that the NIOSH approach could lead to the 

underestimation of ingestion intake for some workers. 

6.1.3 Objective 3: Adequacy of Data 

(1) There are numerous assumptions in parametric values feeding into missed dose 

estimations (both internal and external) made in the site profile that are either not 

supported or not adequately supported by explanation, available data, technical study, or 

references. Many of these parametric assumptions are made seemingly arbitrarily 

without adequate technical bases. In some cases, a value is selected as the assumption 

from a range of rough estimated values to bound a dose parameter that is not entirely 

justified or explained in the document. In other cases, the value selected does not appear 

to be bounding. This is a flaw that could adversely affect the credibility and validity of 

the assigned missed dose estimates in the site profile. 

 

(2) There are significant uncertainties in historical information and data regarding Linde 

facilities, operations, timelines, and practices. The site profile identifies only seven 

buildings at Linde that were used for the MED uranium processing project. However, it 

is not possible to clearly verify the existence of Building A and Building B during the 

MED operation period and, notwithstanding, to identify their MED-related functions. 
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During the site expert interviews, however, many past Linde workers identified other 

buildings that had been involved directly or indirectly with the uranium processing 

operations. These buildings include Buildings 1, 2, 19, 39, 52, 54, 57, and 59. There are 

also outside areas used for staging uranium ore or waste piles that are adjacent to 

Buildings 30, 75, and 76. In addition, NIOSH makes a large number of facility 

operation-related and functional assumptions in this site profile in order to provide a 

framework for dose reconstruction. 

(3) The air concentration data used in the site profile do not appear to be claimant favorable, 

due to significant expert criticism regarding using air concentration data to estimate 

worker inhalation intakes at uranium processing facilities. Several technical studies, 

including the 2003 Y-12 study, Practical Use of Personal Air Sampling (PAS) Data in 

the Internal Dosimetry Program at the Y-12 National Security Complex (Snapp 2003), 

and the Nuclear Regulatory Commission’s NUREG-1400, Air Sampling in the Workplace

(Hickey 1993), demonstrate that using air concentration data could lead to 

underestimating the worker intakes and, subsequently, the internal exposures. The Y-12 

study shows as much as a factor of 10 difference between intakes derived from bioassay 

data and intakes derived from air concentration data (with the bioassay data producing 

the higher doses). In addition, the sensitivity of survey instruments, locations of the air 

samples, and air flow studies of the buildings are not presented in the Linde Site Profile 

report. It is also important to point out that the air concentration data used are based on 

results of random grab-air samples in general areas and breathing zones, and not the 

results of continuous area sampling. Lack of more complete air sampling data and 

neglecting bioassay data could lead to significant uncertainties in worker intakes and 

underestimation of missed inhalation doses. 

(4) Using air concentration data only, but neglecting urinalysis data, to estimate worker 

inhalation intakes in the Linde Site Profile is not in full compliance with 42 CFR 82 

requirements. There are 8 sets of urinalysis data for over 100 uranium workers in the 

ORAU database for the period between December 16, 1947, and January 30, 1950. The 

air concentration data used in the site profile also are not complete and are deemed 

inadequate (see Finding 2). However, NIOSH and ORAU decided to use only these air 

concentration data for dose reconstruction. This approach is not fully compliant with the 

hierarchy approach stipulated in 42 CFR 82. 

(5) The Linde Site Profile uses a very complex scheme of surrogate data to estimate missed 

occupational external dose to workers from 1942 to the present. NIOSH uses a 

combination of film badge data, solid sample analysis results, and facility field 

measurements to fill exposure data gaps in order to cover the entire period of the Linde 

operations. These surrogate data are, however, limited and, most importantly, not 

facility- or building-specific. In many cases, the site profile uses different sets of data to 

estimate worker beta and gamma doses separately. Even though NIOSH attempts to use 

partially conservative assumptions in portions of the external dose estimation, the overall 

approach is questionable and potentially flawed, and may lead to assigned missed internal 

doses that are not claimant favorable. 
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(6) SC&A also found that the site profile does not fully develop procedures and guidelines 

for dose reconstructors for gaps in environmental, internal, and external doses that could 

lead to a significant underestimation of worker dose. Input from interviewed site experts 

indicates that there were situations where reactor workers were not provided neutron 

dosimeters or were not monitored on a continual basis, and where processing facility 

workers were not monitored when they had positive nose smears. Data either presented 

in the site profile, or on which the site profile is based, cannot be considered adequate 

unless an evaluation is conducted of the comprehensiveness of the neutron-monitoring 

and bioassay programs, and of what extent existing dose estimation assumptions and 

methodologies address this potential missed dose. 

(7) The statistical approach used in the Linde Site Profile may not be claimant favorable. In 

Tables 6, 13, 14, 15, 17, 18, 21, 23, 24, 25, 26, 29, 30, 31, 32, and 33, the site profile lists 

the GM or the GSD values for various assigned default assumptions. First, there are no 

supporting calculations or data to show how these geometrical quantities were calculated. 

Second, the use of geometric means and standard deviations of airborne radon 

concentrations (for example) as default values could be considered claimant-neutral and 

not claimant-favorable. Unless there is good reason to believe that a given worker was 

exposed to the full distribution of the measured concentrations and could not have 

experienced protracted exposures to higher than average radon concentrations, it may be 

more appropriate to use the upper 95th percentile as the default exposure level. NIOSH’s 

use of the GSD approach may not address very high, short-term, episodic exposures; 

short-term exposure during incidents; and radon intakes during the performance of tasks 

with a potential for high transient air concentrations. 

6.1.4 Objective 4: Consistency Among Site Profiles 

Selected processes and assumptions contained in the Linde Site Profile are compared with those 

of MCW, Bethlehem Steel, Y-12, INL, SRS, and Hanford site profiles. Mallinckrodt, Bethlehem 

Steel, and Y-12 have some attributes of an AWE and some attributes of a larger DOE facility, 

which made comparisons with DOE and AWE facilities appropriate. 

There are a number of key elements in the Linde Site Profile that are similar to elements 

commonly used by NIOSH in other site profiles. These include a site description, and 

occupational medical dose, internal dose, and external dose discussions. Occupational medical 

dose represents dose from x-rays required as a condition of employment and is usually tracked 

via site medical records. Internal dose represents dose from radioactive material inhaled or 

ingested into the body, including that resulting from residual contamination. External dose 

represents exposure to sources of radiation external to the body from beta particles, photons, 

neutrons, and other particles. However, one large difference between the Linde Site Profile and 

other site profiles is the omission of the Occupational Environmental Dose section in the former, 

which accounts for doses resulting from exposure to radioactive material that has been released 

to the air, water, or soil, or from exposure to waste piles or contaminated soil or ground. 

Table 6-2 contains a summary of the key site profile elements addressed in each of the site 

profiles. 
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Table 6-2: Site Profile Key Element Comparison 

Element Linde MCW Bethlehem Y-12 INL SRS Hanford 

Site Description Included Included Included Included Included Included Included 

Medical Dose Included Included Included Included Included Included Included 

Internal Dose Included Included Included Included Included Included Included 

External Dose Included Included Included Included Included Included Included 

Environmental Dose Not Included Not Included Not Included Included Included Included Included 

(1) Although Y-12, INL, Hanford, and SRS have sections on the development and 

assignment of environmental dose, this is absent from the MCW and Bethlehem Steel site 

profiles, as well as the Linde Site Profile. In the case of Linde, there are available 

information and data indicating the presence of soil and ground contamination in 

different locations of the Linde Site. Section 2.3.4, Other Radiological Activities, of the 

Linde Site Profile states the following: 

Early on, radioactive liquid wastes were discharged to the Tonawanda 

sanitary sewage system. Due to the nature of the liquids, this became a 

problem, and Linde began to dispose of liquid wastes into onsite wells that 

sometimes overflowed. Later still, liquid wastes were discharged to a 

drainage ditch that led to a sewer conduit (BNI 1993, pp. 1-9 to 1-15). 

(Davidson 2005, p. 21)

After MED work began at the Tonawanda site, there was potential 

exposure of workers to radiation and radioactivity when outdoors. 

Portions of the site are known to have been contaminated with 

radioactivity (Heatherton 1948h; ORNL 1978; BNI 1982); resuspension 

would have produced airborne radioactivity. One source of ground 

contamination and airborne radioactivity was the ore unloading process, 

which involved transporting ore in buggies — sometimes in bulk and 

sometimes in bags - from box cars to Building 30 (see the section on Step I 

operations in Section 2.3.2.1). Outdoor areas of the site were sometimes 

used for storage of radioactive materials. Olevitch (1944) reports the 

outdoor storage of contaminated ore bags that at times numbered in the 

thousands. In 1948, 1 mR/h gamma and 3,000 a dpm/100 cm2

 were 

measured from the soil in an area formerly used for storage of radioactive 

materials (Heatherton 1948h). 

An additional source was the release of liquid effluents either to onsite 

wells that sometimes overflowed, or to an onsite drainage ditch (BNI 1993, 

pp. 1-9 to 1-15). Airborne effluents from the plant were an additional 

source of outdoor radioactivity. 
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In 1949, Linde workers were used to unload drums of K-65 shipped by rail 

to the Lake Ontario Ordnance Works at Modeltown, New York (Wolf 1949; 

Heatherton 1949a). Linde film badges were worn during this work. 

At the April 18, 2005, Worker Outreach Meeting, strontium was 

mentioned. Summary information from the meeting did not indicate if this 

was radioactive on nonradioactive strontium, or if this was process 

material or a sealed source. Because no mention of radioactive strontium 

was found in the available Linde documents, it is believed likely that this 

source might have been a sealed radioactive source, nonradioactive 

material or a small quantity in comparison to the uranium source term. 

The estimates of uranium exposure based on estimates of exposure periods 

and source term, which were based on worst case assumptions when a 

parameter was not well supported by available information would be 

sufficiently bounding to account for small amounts of radioactive 

strontium. (Davidson 2005, p. 22) 

In addition, Section 2.7, Post MED/AEC Operations, of the Linde Site Profile states the 

following: 

The Formerly Utilized Sites Remedial Action Program (FUSRAP) began 

in 1976. Oak Ridge National Laboratory (ORNL 1978) surveyed the 

Linde Tonawanda from October 18 through November 5, 1976 to 

determine if remediation would be required. Radiation and radioactive 

contamination measurements were made inside Buildings 14, 30, 31, 37 

and 38; on the Tonawanda property outside the buildings; and at nearby 

offsite locations. Linde employees noted that Building 30 renovation 

occurred in the 1960s and could have resulted in elevated employee 

radiation exposures. Notes from the Worker Outreach Meeting on April 

18, 2005 mention contamination associated with Building 57, and an 

additional review of the BNI 1993 remedial investigation report shows 

areas of residual radioactive contamination were associated with areas in 

or near Buildings 57, 58 and 90. The highest indoor radiation levels were 

found in the principal production buildings, 30 and 38. Linde was 

designated as a FUSRAP site in 1980. Additional radiological surveys 

and decontamination efforts followed (BNI 1993). These led eventually to 

demolition of Building 14 and all of the Linde Ceramics buildings 

involved in MED/AEC work except Building 31. Table 1 shows demolition 

dates. As of 2004, Building 31 remained in use, and onsite soil 

remediation was in progress with completion scheduled for 2007 

(Pilon 2004). 

Based on post-operational concentrations of radioactive material in the soil and water, it 

is reasonable to expect similar or greater levels of environmental release during the 

production years. This is further collaborated by site expert interviews, stating that 

uranium dust and fumes were released from production buildings. The likelihood of 
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exposure from environmental release is significant and requires further consideration in 

the site profile. Also, for consistency and completeness of the site profile, environmental 

exposure should be addressed, whether or not the dose consequences are judged to be 

significant. 

(2) The assignment of dose for occupational medical exposure is discussed in ORAUT-TIB0006 (Kathren 2003), which is applicable to both AWE and DOE facilities. In the cases 

of Y-12, INL, Hanford, and SRS, there were considerable data on the particular x-ray 

units and medical monitoring procedures. Due to the lack of medical records at Linde, 

MCW, and Bethlehem Steel, an annual standard chest x-ray was assumed for each 

covered year of the facility. The SRS and Hanford TBDs assumed early x-rays included 

photofluorography (PFG), while the Linde and MCW site profiles did not include PFG. 

Photofluorography was a common technique prior to 1960 (OCAS 2002) and should be 

considered unless solid evidence to the contrary can be provided. Additionally, the 

MCW site profile utilizes different analogue organs for input into IREP than the SRS 

TBD. In the case of default assumptions, such as kVp, mAs, source-to-image distance, 

and uncertainty, the approach is consistent among the site profiles examined. 

(3) The methodology for assignment of internal dose is somewhat specific to each site. In 

the case of Linde, MCW, Y-12, INL, SRS, and Hanford, urinalysis data were available 

for a portion of the covered period. This information was used as the primary source for 

internal dose calculation for MCW, Y-12, INL, SRS, and Hanford, supplemented with air 

concentration data. Instead, the Linde Site Profile uses only airborne uranium dust 

concentration data as the primary source for missed occupational internal dose 

calculation. Urinalysis data is used as a predictive comparison only. Other issues of 

concern in internal dose assumptions between Linde and the other site profiles include 

the following: 

• Linde, Bethlehem Steel, and MCW site profiles have statistical procedures and or 

data tables that result in dose estimates that do not give claimants the benefit of the 

doubt in the face of uncertainties in at least some cases. 

• The Linde Site Profile uses inconsistent work hours, while the Bethlehem Steel Site 

Profile assumes 10-hour workdays, and the MCW Site Profile assumes 2,000 hours 

per year (effectively, an 8-hour workday.) 

• The Linde Site Profile assumes a factor of 0.2 of inhalation intake for ingestion intake, 

while an explicit analysis of ingestion dose is included in the Bethlehem Steel Site 

Profile (Revision 01), and ingestion dose was deemed to be negligible in the MCW 

Site Profile. Large particle ingestion was not taken into account in any of the three 

cases. 

(4) In the case of Linde, MCW, Y-12, INL, SRS, and Hanford, film badge data is used as a 

primary source of external exposure, where available. Some inconsistencies in external 

dose assumptions between Linde and the other site profiles include the following: 
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• Adjustment factors are applied to recorded external dose at SRS and Hanford to 

estimate Hp(1.0) doses. The Linde Site Profile does not recommend any adjustment 

to recorded dose. The site profile, therefore, does not take into consideration the 

laboratory, radiological, and environmental uncertainties in the personal dosimetry 

program. 

• Due to the limited nature of the film badge data, the Linde Site Profile supplements 

the external dose calculations with survey measurements. However, the site profile 

did not take into consideration the radiological and environmental uncertainty 

associated with the survey instruments. Therefore, no adjustment factors are 

recommended to the measurements or the estimated missed external dose. 

• The Linde Site Profile assumes various numbers of work hours per week, while the 

Bethlehem Site Profile assumes 10-hour workdays and the MCW Site Profile 

assumes 2,000 hours per year (effectively an 8-hour workday.) 

(5) SC&A recognizes that operations, exposure conditions, and facility designs vary, even in 

facilities with the same production mission. We have presented some comparisons of the 

Linde Site Profile with other site profiles. Based on the source term and the workplace 

conditions at the Linde Site, variations in assumptions may be justified. Further 

explanation in the Linde Site Profile, however, is required to account for the alternate 

assumptions versus other site profiles in light of the similarities in source terms at MCW, 

Y-12, portions of Hanford, and portions of the SRS. 

6.1.5 Objective 5: Regulatory Compliance 

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

implementation guides. However, SC&A has identified some significant shortcomings of the 

data used in the review process of the Linde Site Profile that may lead to dose reconstructions 

that are not claimant favorable. It is especially crucial for NIOSH to re-evaluate the technical 

basis of the missed dose assumptions: 

(1) Using air concentration data, but neglecting urinalysis data, to estimate worker inhalation 

intakes in the Linde Site Profile is not in full compliance with 42 CFR 82 requirements. 

There are 8 sets of urinalysis data for over 100 uranium workers in the ORAU database 

for the period between December 16, 1947, and January 30, 1950. The air concentration 

data used in the Linde Site Profile are not complete and are deemed inadequate (see 

Finding 2). However, NIOSH uses these air concentration data only for dose 

reconstruction. This approach is not in full compliance with the hierarchy approach 

stipulated in 42 CFR 82. 

(2) An assessment of uncertainties, as required by OCAS-IG-001 and OCAS-IG-002, has not 

been adequately developed for air concentration and radon measurement data used in lieu 

of bioassay data. As characterized in the site profile, “little information was available” 

during the periods of production and non-production. In fact, the site profile uses 

different words to indicate that the information gathered is “uncertain,” such as 
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“probably,” “likely,” and “assumes.” It gives the strong impression that the available 

data gathered are inaccurate and uncertain. NIOSH should develop a method to 

determine best estimates and their uncertainties, as well as upper 95th percentile values of 

time-weighted inhalation intake, radon intake, beta dose, gamma dose, and neutron dose, 

for both internal and external dose calculations. 

6.2 USABILITY OF SITE PROFILE FOR INTENDED PURPOSE 

SC&A has identified seven criteria that reflect the intent of the EEOICPA and the regulatory 

requirements of 42 CFR Part 82 for dose reconstruction. Because the purpose of a site profile is 

to support the dose reconstruction process, it is critical that the site profile assumptions, analytic 

approaches, and procedural directions be clear, accurate, complete, and auditable 

(i.e., sufficiently documented). SC&A used the following seven objectives to guide its review of 

the Linde Site Profile to determine whether it meets these criteria: 

• Objective 1 − Determine the degree to which procedures support a process that is 

expeditious and timely for dose reconstruction 

• Objective 2 − Determine whether procedures provide adequate guidance to be efficient in 

select instances where a more detailed approach to dose reconstruction would not affect 

the outcome 

• Objective 3 − Assess the extent to which procedures account for all potential exposures, 

and ensure that resultant doses are complete and are based on adequate data 

• Objective 4 − Assess procedures for providing a consistent approach to dose 

reconstruction, regardless of claimants’ exposures by time and employment locations 

• Objective 5 − Evaluate procedures with regard to fairness and the extent to which the 

claimant is given the benefit of the doubt when there are unknowns and uncertainties 

concerning radiation exposures 

• Objective 6 − Evaluate procedures for their approach to quantifying the uncertainty 

distribution of annual dose estimates that is consistent with and supports a Department of 

Labor probability of causation estimate at the upper 99% confidence level 

• Objective 7 − Assess the scientific and technical quality of methods and guidance 

contained in procedures to ensure that they reflect the proper balance between 

current/consensus scientific methods and dose reconstruction efficiency 

The following items address these objectives: 

(1) The Linde Site Profile does not meet the objectives in the Occupational Medical Dose 

section in assessing the potential organ doses to workers who received medical x-ray 

examinations over the entire operating history of the Linde Site. However, the Linde Site 
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Profile is adequate to instruct the dose reconstructors to use the appropriate TIB for the 

estimation of missed occupational medical dose to workers. 

(2) The Linde Site Profile does not meet the objectives in assessing the potential 

occupational environmental doses to Linde workers. In fact, the Linde Site Profile does 

not present any environmental exposure data. 

(3) The Linde Occupational Internal Dose section does not use available, though limited, 

urinalysis data; instead, it uses only airborne uranium dust concentration data for the 

estimate of missed occupational internal dose to workers. 

(4) The Occupational Internal Dose section uses very limited surrogate data from 1944 for 

the internal dose calculations for the entire period of 1942 to 1949 that are not fully 

supported. 

(5) The Occupational External Dose section uses a very complex surrogate data system to 

estimate missed occupational external dose to workers for the entire period. 

(6) The site profile does not meet the objective in assessing potential occupational 

environmental dose, because it neglects evaluating doses from burlap bags, contaminated 

soil, contaminated rooftops, contaminated underground tunnels, contaminated injection 

wells, waste tanks, waste piles, and other potential airborne sources. 

(7) The site profile makes a number of assumptions that are not claimant favorable, and in 

many cases, are either not supported or fully supported; it does not meet the objectives. 

6.3 UNRESOLVED POLICY OR GENERIC TECHNICAL ISSUES 

A number of issues are identified that are common to the Linde, MCW, Bethlehem Steel, Y-12, 

Rocky Flats, INL, Hanford, and SRS site profiles and, in some cases, represent potential generic 

policy issues that transcend any individual site profile. These issues may involve the 

interpretation of existing standards, how certain critical worker populations should be profiled 

for historic radiation exposure (e.g., construction workers and early workers), and how exposure 

itself should be analyzed (e.g., treatment of incidents and statistical treatment of dose 

distributions). NIOSH indicates that it may develop separate TIBs in order to address some of 

these generic issues. The following presents those issues identified in the Linde Site Profile 

Review that SC&A believes represent transcendent issues that need to be considered by NIOSH 

as unresolved policy or generic technical issues. 

(1) Direction on the applicability and usability of the site profiles, TBDs, and/or TIBs to 

individual dose reconstructions is absent. 

(2) Examples of dose reconstruction are seldom provided in the site profiles, TBDs, and/or 

TIBs. Examples are extremely useful for dose reconstructors to follow and use the 

guidance accurately. 
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(3) Adequacy and completeness of worker records are essential to claimant-favorable dose 

reconstructions. None of the site profiles address this issue or give direction on resolving 

issues with missing records. 

(4) Site expert testimony indicates that many workers moved from one plant to the next on 

the same site, creating a complicating factor in determining overall exposure. 

Establishment of an accurate worker history is crucial in such cases. This is especially 

difficult to accomplish in cases of family-member claimants, where the survivors cannot 

be expected to have a good grasp of where and when the worker was stationed. 

(5) Statistical techniques used in the application of the data to individual workers should be 

considered. However, using statistical averages may not be claimant favorable, since in 

most compensable cases, they would not provide the upper bound for missed worker 

doses. For example, in the Linde Site Profile, GM are used to estimate the missed worker 

doses instead of a more appropriate and claimant-favorable upper 95th percentile value. 

(6) Dose from impurities and/or daughter products in radioactive material received and 

processed at sites should be assessed. 

(7) Assumptions on solubility, breathing rate, and ingestion should be addressed. 

(8) A correction factor for external gamma doses should be considered to account for angular 

dependence of dosimeter sensitivities. 

(9) Direction with respect to consideration of incidents and high-risk (dose) jobs in 

individual dose reconstructions should be provided. In the case of the Linde Site Profile, 

high-risk job categories are much better defined and classified than in other site profiles 

or TBDs. 

(10) Availability of monitoring records for subcontractor and/or visitor personnel while 

working on or visiting a facility should be ascertained. In the case of the Linde Site 

Profile, no discussion was made on this subject. 

(11) Dose to construction workers and other early workers should be assessed. In the case of 

the Linde Site Profile, no discussion was made on this subject. 

(12) Unique exposure conditions for cleanup and decontamination and decommissioning 

workers should be considered. The relative impact of each of these items on dose 

reconstruction is site-specific and requires independent evaluation in each site profile or 

TBD. 
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7.0 REFERENCES 

42 CFR Part 81, 2002. Title 42, Public Health, Part 81, Guidelines for Determining the 

Probability of Causation Under the Energy Employees Occupational Illness Compensation 

Program Act of 2000; Final Rule, Code of Federal Regulations, May 2, 2002.

42 CFR Part 82, 2002. Title 42, Public Health, Part 82, Methods for Radiation Dose 

Reconstruction Under the Energy Employees Occupational Illness Compensation Program Act 

of 2000, Code of Federal Regulations, May 2, 2002. 

ACE 2000, Record of Decision for the Linde Site, Tonawanda, New York, March 2000, U.S. 

Army Corps of Engineers, Buffalo District. 

AEC 1949a, “Health Hazards in NYOO Facilities Producing and Processing Uranium: A Status 

Report- April 1, 1949,” Prepared by the New York Operations Office Medical Division, Issued 

April 18 [Mallinckrodt\08-28-03 – Data Capture\010001510]. 

ACE Buffalo (U.S. Army Corps of Engineers Buffalo District), 2003, Record of Decision for the 

Building 14 Operable Unit, Linde Site, Tonawanda, New York, April. [Linde Air and Ceramics, 

NY\12-06-04- Data Capture\030001716.pdf Linde Air and Ceramics, NY\12-06-04- Data 

Capture\030001719.pdf] 

ACE Buffalo (U.S. Army Corps of Engineers Buffalo District), 2004b, Air Monitoring Data, 

Linde Remedial Action Project, Linde FUSRAP Site. [Linde Air and Ceramics, NY\12-06-04- 

Data Capture\030001726.pdf] 

Allen, D., 2002, Internal Dose Reconstruction Implementation Guidelines, Revision 0, OCASIG-002, NIOSH, Cincinnati, Ohio, August, 2002. 

Aerospace 1981, Evaluation of the 1943-1946 Liquid Effluent Discharge from the Linde Air 

Products Company Ceramics Plant, Aerospace Corporation, Germantown, Maryland, December. 

Blatz, H., 1949, “Report on a General Radiation Survey at Linde Air Products Company, Div. of 

Union Carbide & Carbon Corp, Ceramics Plant, Tonawanda,” U.S. Government office 

memorandum, April 19. [Linde Air and Ceramics, NY\07-22-03 - Data Capture 2-

Linde\Eml2Box25LindeAirProducts46_50.pdf, p. 18]. 

BNI 1982, Preliminary Engineering Evaluation of Remedial Action Alternatives, Linde Air 

Products Site, Tonawanda, New York, Formerly Utilized Sites Remedial Action Program 

(FUSRAP), Contract No. DE-AC05-810R20722, Bechtel Job 14501, BNI-010-05-29-59A, 

November. [Linde Ceramics, NY\12-06-04-Data Capture\030001718.pdf] 

BNI 1993, Remedial Investigation for the Tonawanda Site, DOE/OR21949-300, Oak Ridge, 

Tennessee. 
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Davidson, G.R., 2005, An Exposure Matrix for Linde Ceramics Plant (including Tonawanda 

Laboratory), ORAUT-TKBS-0025, Revision 0, Office of Compensation Analysis and Support, 

Cincinnati, Ohio, May 31, 2005. 

Dupree, E. A., 1983, "Linde Ceramics Plant: Interview with (name withheld) and (name 

withheld)," AU0004849, memorandum to Linde Ceramics Plant File, Oak Ridge Associated 

Universities, December 2 [Linde Air and Ceramics, NY\08-07-03 - Data Capture\010001224 

p. 2]. 

Ferry, J. L., 1944b, "Radiation Hazards," EIDMM-3a, memorandum to Area Engineer, 

Tonawanda Area (Tonawanda, NY), February 29 [Linde Air and Ceramics, NY\10-20-03 – Data 

Capture\240000404, p. 3]. 

Heatherton, R. C., 1948h, “Progress Report,” Linde Ceramics Plant, December 20 [Linde Air 

and Ceramics, NY\05-25-04 - Data Capture\250000092.pdf, p. 54]. 

Heatherton, R. C., 1949a, "Unloading K-65 from Railroad Car at Modeltown," letter to B. S. 

Wolf (Medical Section, New York office, U.S. Atomic Energy Commission), Linde Air Products 

Company, Ceramics Plant, Kenmore, New York, March 25 [Linde Air and Ceramics, NY\08-07-

03 -Data Capture\010001228\000002 p. 1]. 

Heatherton, R. C. 1950, Decontamination and Survey of Uranium Refinery Plant, U.S. Atomic 

EnergyCommission, NYO-1536, December 14 [Linde Air and Ceramics, NY\Linde Air 

Products.pdf p.133]. 

Hickey 1993, “NUREG-1400, Air Sampling in the Workplace”, Final Report, Nuclear 

Regulatory Commission, September 1993. 

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract 

Modelfor Radiological Protection, ICRP Publication 66, Pergamon Press, Oxford, England. 

Kathren, R. L., 2003, Technical Information Bulletin: Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Projections, ORAU-OTIB-0006, Rev.00, November 14, 2003. 

Linde 2004a, Linde Ceramics Plant Urinalysis Data, 1947–1950, miscellaneous reports of 

uranium and fluorine urinalysis data for Linde Ceramics' workers. Linde 2004b004a; Linde 

Ceramics, NY\03-09-04 – Data Capture\260000120, pp. 1–29. 

Linde 2004b, Linde Ceramics Plant Urinalysis Data, 1947–1950, miscellaneous reports of 

uranium and fluorine urinalysis data for Linde Ceramics' workers. Linde 2004b004b; Linde 

Ceramics, NY\01-22-04 – Data Capture\260000009, pp. 1–48. 

Merwin, S. E., 2005, Assignment of Missed Neutron Doses Based on Dosimeter Records, 

ORAUT-OTIB-0023, Revision 00, Oak Ridge Associated Universities, Oak Ridge, Tennessee, 

March 7, 2005. 
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NIOSH 2004, Estimation of Ingestion Intakes, OCAS-OTIB-009, Revision 0, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, April 13, 2004. 

Olevitch, E., 1944, Processing L-30 Ore in Step I, internal report, Linde Ceramics Plant, July 15 

[Linde Air and Ceramics, NY\08-07-03 - Data Capture\010001220\000003 p. 40]. 

 

ORNL 1978, Radiological Survey of the Former Linde Site, Tonawanda, New York, February 

1978. 

Pilon, R. L., 2004, untitled attachment to "RE: Request or Information on Linde Site Cleanup 

and Remediation," e-mail to Gerald R. Davidson, from U.S. Army Engineer District – Buffalo, 

New York, October 8. [Linde Air and Ceramics, NY\12-06-04- Data Capture\030001725.pdf] 

Rollins, G., 2004, Technical Information Bulletin – Maximizing Internal Dose Estimates for 

Certain DOE Complex Claims, ORAUT-OTIB-0002, Revision 01 PC-2, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, May 7, 2004. 

SC&A, 2004a. SC&A Standard Operating Procedure for Performing Site Profile Reviews, 

Rev. 0, March 18, 2004. 

Snapp, L, M. 2003, Practical Use of Personal Air Sampling (PAS) Data in the Internal 

Dosimetry Program at the Y-12 National Security Complex, RCO-2003-004, Oak Ridge, 

Tennessee. 

Strom, D., 2003. Technical Information Bulletin – Occupational Dose from Elevated Ambient 

Levels of External Radiation, ORAUT-OTIB-0007, Revision 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, November 12, 2003. 

Taulbee, T., 2002. External Dose Reconstruction Implementation Guidelines, Revision 1, 

OCAS-IG-001, Office of Compensation Analysis and Support, NIOSH, Cincinnati, Ohio, 

August 2002. 

Wolf, B. S., 1949, “Linde Air Products Co., Division of Union Carbide & Carbon Corp, 

Ceramics Plant, Tonawanda, N.Y. – Decontamination of Main Building,” memorandum to F. M. 

Belmore, Production Division, from Medical Section, New York office, U.S. Atomic Energy 

Commission, May 12 [Linde Air and Ceramics, NY\07-22-03 - Data Capture 2-

Linde\Eml2Box25LindeAirProducts47_52.pdf, p. 13]. 
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ATTACHMENT 1: CONFERENCE CALL WITH NIOSH AND SC&A 

SC&A Questions/Comments for NIOSH 

Regarding Linde Site Profile 

SC&A’s evaluation of the Linde site profile (the “Site Profile”), An Exposure Matrix for Linde 

Ceramics Plant (including Tonawanda Laboratory) (ORAUT-TKBS-0025, Rev. 00, 5/31/05, 

Davidson 2005) reviewed reference documents, source records, and interviews with former 

Linde workers and retirees. The documents include primarily sources obtained from the 

FUSRAP site, declassified documents regarding early worker exposures, and a majority of the 

key reference and source documents contained in the NIOSH/ORAU Site Research Database. In 

addition to document review, Kathy Robertson-Demers and Desmond Chan conducted 

interviews with 17 Linde site experts in Buffalo, New York, on January 12–14, 2006. The 

workers discussed the functions and operations they individually performed at the various Linde 

buildings and facilities that made up the Linde Ceramics Plant. As a result of the document 

reviews and interviews with the workers, a number of issues with respect to the Site Profile have 

surfaced, and SC&A has formulated the following questions to help inform its Linde Site Profile 

review. For convenience and clarity, the questions have been organized into sections of related 

issues. 

General Questions 

1. The Site Profile estimates airborne concentrations between October 1, 1942, and 

December 31, 1954, for the Linde Ceramics Plant in Table 9 (p. 34); internal exposures to 

workers at the Ceramics Plant between October 1, 1942, and December 31, 1954, in 

Table 11 (p. 37); and internal exposures to workers at the Tonawanda Lab between 

October 1, 1942, and December 31, 1954, in Table 12 (p. 38). However, the cells for the 

period between 1942 and 1947 are designated as “reserved.” Please clarify what is meant 

by this designation; will these data be provided in a future revision of the Site Profile? 

2. The Linde workers that SC&A interviewed indicated that their employment and medical 

records were shipped to Union Carbide’s record repository in Vermont. Has 

NIOSH/ORAU visited this repository to review these records? If a visit was conducted 

and the records evaluated, what do the records contain? Are they sufficient to support 

individual claimant dose reconstruction? Were they used to develop the Site Profile? 

3. The Site Profile mentions that there were episodic events at Linde and Tonawanda during 

the production and cleanup periods. Did NIOSH/ORAU construct a list of incidents or 

determine the frequency of episodic problems? Did NIOSH/ORAU determine the 

potential exposure to workers who were present or participated in the activities? 

4. On October 12, 2005, NIOSH issued an evaluation report on the Special Exposure Cohort 

(SEC) Petition 00044, submitted on September 29, 2005, covering a class of atomic 

weapons employees at the Linde Ceramics Plant from October 1, 1942, to October 31, 
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1947. In this report, NIOSH concludes that dose reconstruction could not be completed 

because of a lack of sufficient information. 

Members of this class at the Linde Ceramics Plant may have received 

internal and external radiation exposures from the uranium and uranium 

progeny in the ores received and processed at the plant. NIOSH lacks any 

biological monitoring data or sufficient air monitoring information or 

sufficient process and radiological source information to estimate the 

potential airborne concentrations to which the proposed class may have 

been exposed (i.e., internal exposures).

What guidance would NIOSH/ORAU provide to dose reconstructors on evaluating nonpresumptive cancers, such as prostate cancers and skin cancers? How should a dose 

reconstructor evaluate worker external exposures for the period between October 1, 1942, 

and October 31, 1947? How could a dose reconstructor separate external exposures from 

internal exposures during this period? Is NIOSH going to issue a revision of the Site 

Profile taking into account the SEC conclusions? 

Site Description and Operational History 

1. Personnel interviewed by SC&A indicated that there was an extensive underground 

tunnel system connecting many of the buildings at the Linde Site that was frequently 

flooded by rain or melting snow, and contaminated with radioactivity. Did 

NIOSH/ORAU take into account potential missing exposures to workers in these tunnels? 

2. Personnel interviewed by SC&A indicated that several MED (Manhattan Engineering 

District)-era buildings at the Linde Site were wooden with cinder blocks, and that the 

wooden structures could not be decontaminated. Did NIOSH/ORAU take this fact into 

consideration when estimating potential exposures to workers after the MED period? 

3. The 2000 Army Corps of Engineers Record of Decision for Linde indicates that there 

were 8,000 tons of filter cakes sent to Ashland, 20,500 tons of waste material sent to 

LOOW, and other material sent to duPont and other facilities. Where were these material 

stored and staged in Linde before they were transported off the site? How were they 

controlled? How were they transported? The Mallinckrodt Site Profile indicates that 

material shipped from Linde was transported in 55-gallon drums. Were similar drums 

also used in shipping filter cakes? The Linde Site Profile does not provide information or 

exposure estimates for unmonitored workers outside the identified production buildings 

in moving, loading, and transporting these waste materials; has NIOSH/ORAU 

considered missed doses to the unmonitored workers involved in transporting these waste 

materials? Similarly, page 10 of the Site Profile states: “The Ceramics Plant produced 

uranium materials for the MED and the AEC from 1943 to 1946 and from 1947 to 1949. 

In the 1947 to 1949 period (and perhaps earlier), Linde received UO2 for processing from 

Mallinckrodt Chemical Works (AEC 1949a).” How was the UO2 transported to the 

production facilities at Linde and then unloaded? Has NIOSH/ORAU considered missed 

doses to unmonitored workers involved in moving, loading, unloading, and transporting 

these uranium products? 
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4. On page 14 of the Site Profile, the description of Building 31 (Ceramics Plant) states, “A 

partially-readable 1945 plant drawing indicates that it contained...a small number of 

associated offices (Linde Database 1945).” Based on the former and retired worker 

interviews, many of the MED production buildings consisted of administrative and 

clerical offices right next to the production areas. There were no separated ventilation 

systems, engineering barriers, or separated entrance or egress in the building for these 

administrative offices. Production workers and administrative staff could move from 

offices to production areas (or vice versa) without any monitoring or radiological control. 

Has NIOSH/ORAU considered potential missed internal and external doses to the 

administrative workers from the production operations? In the Linde Site Profile for the 

standby period from 1946 to 1947, the gamma and beta radiation rates in an office 

building were assumed to be zero, leading one to believe that NIOSH/ORAU did not 

consider potential missed doses to office workers from potential significant cross 

contamination. Please confirm and explain the basis of this assumption. 

5. During the former and retired worker interviews, the workers indicated that the 

contaminated piping in Building 30 and other buildings were not removed and were, 

instead, reused for new production and operations. Did NIOSH/ORAU consider the 

potential missed exposures to workers from previously contaminated piping in 

subsequent activities? 

Occupational Environmental Doses 

1. The Linde Site Profile does not address potential occupational environmental dose to 

workers in a separate section. However, SC&A found many places in the Site Profile 

indicating the potential for such exposure from outdoor, airborne radon or uranium dust, 

contaminated material, soil, or waste piles. Examples include: 

After MED work began at the Tonawanda site, there was potential 

exposure of workers to radiation and radioactivity when outdoors. 

Portions of the site are known to have been contaminated with 

radioactivity (Heatherton 1948h; ORNL 1978; BNI 1982); resuspension 

would have produced airborne radioactivity. One source of ground 

contamination and airborne radioactivity was the ore unloading process, 

which involved transporting ore in buggies — sometimes in bulk and 

sometimes in bags — from box cars to Building 30 ... Outdoor areas of the 

site were sometimes used for storage of radioactive materials. Olevitch 

(1944) reports the outdoor storage of contaminated ore bags that at times 

numbered in the thousands. In 1948, 1 mR/h gamma and 3,000 a dpm/ 

100 cm2 were measured from the soil in an area formerly used for storage 

of radioactive materials (Heatherton 1948h). (p. 22) 

An additional source was the release of liquid effluents either to onsite 

wells that sometimes overflowed, or to an onsite drainage ditch (BNI 1993, 

pp. 1-9 to 1-15). Airborne effluents from the plant were an additional 

source of outdoor radioactivity. (p. 22) 
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The average outdoor beta dose rate to which workers would have been 

exposed during production was assumed to be at most equal to the indoor 

level in Building 30 based on the reasoning in Section 4.1.2.1. Therefore, 

the outdoor beta rate was estimated as 0.676 mrem/h (based on beta dose 

rate at 3’ before vacuum cleaning and flushing given in Table 13). For 

0.5 h/wd exposure, 6 wd/wk, and 50 wk/y, the average worker exposure 

would have been 0.10 rem/y. This is negligible compared to the indoor 

doses given the approximate nature of the estimates and was ignored.

(p. 46) 

The average outdoor gamma exposure rate to which workers would have 

been exposed during production was assumed to be at most equal to the 

indoor level in Building 30 based on the reasoning in Section 4.1.2.1. 

Therefore, the outdoor gamma rate was estimated as 0.131 mR/h based on 

the level before vacuuming and flushing in Table 13. For 0.5 h/wd 

exposure, 6 wd/wk, and 50 wk/y, the average worker exposure would have 

been 0.020 R/y. This is negligible compared to the indoor exposures given 

the approximate nature of the estimates and was ignored. (p. 54) 

In recent years, outdoor air radioactivity has been measured in 

conjunction with the site soil remediation program (ACE Buffalo 2004b). 

Table 38 displays the highest monthly average air concentrations of 

Ra-226, Th-230 and U-238 for July 2000 through June 2004. Assuming 

that the measured U-238 is part of a natural uranium source term, the 

total uranium concentration would be the U-238 concentration divided by 

the fraction of activity due to U-238, 0.4886, or 4.3E-04 pCi/m3

. The 

activity ratios of Th-230 and Ra-226 to uranium are 0.84 and 1.7, 

respectively. (p. 72) 

Since discussions of potential occupational environmental doses are scattered throughout 

the Site Profile, has NIOSH/ORAU provided adequate and comprehensive guidance to 

reviewers to determine doses to workers (monitored and unmonitored), who may have 

been exposed to environmental sources of radioactivity during their outdoor activities 

(loading, unloading, moving, transportation, cleanup)? Specifically, did NIOSH/ORAU 

consider potential occupational environmental doses to unmonitored (non-production) 

workers who might have worked near the MED buildings? 
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2. During the site expert interviews, the workers indicated that there were large ventilators 

in different MED buildings, such as Buildings 14, 30, and 38. These ventilators were not 

equipped with any special filtering systems, and uranium ore and dust were vented out 

directly to the outside atmosphere. Did NIOSH/ORAU consider environmental 

exposures to workers from these exhaust streams? Did NIOSH/ORAU consider 

cumulative contamination on rooftops, external building structures, parking lots, soils, 

and open areas within the Linde perimeter? 

Occupational Internal Doses 

1. Section 3.0 of the Site Profile, Estimation of Internal Exposure, 1942–1954, states: 

To expedite preparation of this document, the Linde information is 

considered in conjunction with information from facilities that did similar 

types of uranium processing to establish preliminary estimates of internal 

intakes and exposures. These estimates are considered best estimates, 

until data can be further considered. It is believed that additional analysis 

of the data will lower at least some of the intakes and exposures estimated 

in this section. (p. 26) 

a. Which are these similar uranium processing facilities that were used in conjunction 

with Linde to establish preliminary estimates of internal intakes and exposures? 

What is meant by “preliminary?” 

b. What similarities did these facilities have to Linde? What information did NIOSH 

and ORAU use from these facilities? Are data from these facilities used as surrogate 

data for Linde? If so, please identify these data and their sources. 

c. What is the definition for best estimates? Are they worst-case estimates, geometrical 

means, or statistical averages? Did NIOSH/ORAU estimate the uncertainties 

associated with these estimates? 

d. What does NIOSH/ORAU mean by “be further considered?” Why does 

NIOSH/ORAU believe that additional analysis of the data would lower the intake and 

exposure estimates? 

2. On page 27 (third full paragraph) of the Site Profile, it is stated, “indications are that 

some of the higher routine (versus episodic) exposures occurred at the uranium ore 

processing facilities.” Can NIOSH/ORAU elaborate on what indications lead to this 

conclusion? How much higher were these routine exposures? Did NIOSH/ORAU 

establish the bounding conditions for these exposures or estimate missed doses in 

reference to these higher routine exposures? 

3. The Linde Site Profile recognizes that African ore involved significantly higher 

exposures to Th-230, Ra-226, and Rn-222. The ingrowth of Th-230 and Pa-231 was not 

considered significant to the production period, but these isotopes may have been 

contributors to external dose during and after the production period. In addition, the 

extraction of uranium would leave a raffinate with a higher concentration of these 

radionuclides, as well as Ra-226. Did NIOSH/ORAU consider the dose contribution of 
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these isotopes? What data or references can NIOSH/ORAU provide to support the 

conclusion that uranium progeny would not contribute to worker doses in this period? 

4. The Site Profile lists Ac-227 and Pa-231 as radionuclides of concern for internal 

exposure to Linde workers up to 1954 in Tables 5 (p. 31), 11 (p. 37), and 12 (p. 38). 

However, these isotopes are not included in Table 39 (p. 74) for worker exposure during 

the subsequent cleanup period; please clarify. 

5. On page 28 of the Site Profile, NIOSH/ORAU assumes air concentrations at 300 MAC 

(maximum allowable air concentration MAC or preferred level PL) adequate to provide a 

quick estimate of exposure even though short term exposures might be higher. What is 

the basis of the 300 MAC value for Linde? What were the potential short-term exposures, 

and what kind of job categories might these exposures relate to? 

On the same page (third paragraph) of the Site Profile, it states: 

Beginning November 1, 1947 at Linde Ceramics, workers were assumed to 

be exposed to 33 MAC and it was assumed this exposure continued 

through cleanup in 1954. Uranium progeny are not included in this later 

period, because only refined uranium was used and because the dose from 

intakes of contamination left from earlier work would have been 

insignificant compared to the dose to uranium during operations. (p. 28) 

What is the basis to assume 33 MAC as the intake concentration for workers from 

November 1, 1947, through December 31, 1954, when earlier on the page the Site Profile 

assumes 300 MAC? The 33 MAC value is a time-averaged statistical mean value and not 

a bounding, worst-case concentration for workers, who might have been involved with 

higher short-term exposures; hence, this value does not appear to be claimant favorable. 

6. Page 28 of the Site Profile states: 

It was initially and arbitrarily assumed that exposures decreased to 

1 MAC during the standby period at the Ceramics Plant, and that 

exposures decreased to 0.1 MAC at the Tonawanda Laboratory after 

cleanup in 1946 until the end of cleanup at the Ceramics Plant in 1954. 

Based on reviews of later air concentrations at Linde, and reviews of air 

concentration data from other sites, it is believed that most workers’ 

exposures would have been much lower during these periods. (p. 28) 

What basis did NIOSH/ORAU use to decrease to 1 and 0.1 MAC? Can NIOSH/ORAU 

provide data to support this assumption and references for the “later air concentrations at 

Linde?” 

7. The Site Profile uses several different workweek hour assumptions for missed dose 

calculations. In many cases, the value of 40 hours per week is used. However, Table 4 

(p. 24) assumes that Linde workers could have worked as much as 51 or 54 hours per 

week. The differences of 11 or 14 hours per week would give significantly greater dose 

values; 27% or 35%, respectively. What basis does NIOSH/ORAU have to conclude that 
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other factors, such as “assumed” air concentrations, are sufficiently large to account for 

these differences, as stated on page 28 (fourth paragraph)? 

To simplify calculations, it assumed that the workweek was 40 hours long 

during all years, although it is likely that the workweek for many was in 

excess of 40 hours especially during the earlier years. The assumed air 

concentrations are sufficiently large to account for any differences in 

actual hours exposed. (p. 28) 

Assuming that one likely, but unquantified, overestimate is sufficient to account for 

another underestimate does not appear conservative: i.e., claimant favorable. 

 

8. Page 35 of the Site Profile states: 

For example, the annual uranium inhalation intake due to chronic 

exposure at 0.1 MAC is estimated by multiplying the air concentration of 

7 dpm/m3

 by the alpha fraction of uranium, 0.402; the ICRP 66 (ICRP 

1994) recommended breathing rate of 1.2 m3/h; and the assumed 2000 

work-hours per calendar year. This results in an annual chronic 

inhalation intake of 6.75E+03 dpm, which is equal to a daily intake rate of 

18.5 dpm/day. For the assumed exposure at 33 MAC, no alpha activity is 

apportioned to progeny so the daily uranium intake would be 

1.52E+04 dpm/day. (p. 35) 

Why did NIOSH/ORAU choose the value of 1.2 m3

/h for Linde, while 1.4 m3/h or 

1.7 m3/h for heavy work was used for the Mallinckrodt Chemical Plant? Similarly, why 

did NIOSH/ORAU use 2,000 work hours per calendar year instead of 2,040 hours (on 

page 32) or, more conservatively, 2,550 or 2,700 hours per calendar year (Table 4)? 

9. Page 29 of the Site Profile states that “individual uranium urinalysis data from November 

1947 to 1950 are available for some Linde Ceramics workers (Linde Ceramics Plant 

Urinalysis Data 1947–1950).” How many workers have these individual urinalysis data 

in their files? How complete are these data? For those workers who have no individual 

urinalysis data, what recommendation is given to the dose reconstructor? 

10. Section 3.2.2, Uranium Air Concentrations, states: 

This site profile currently uses an estimate of Linde’s largest time 

weighted air concentration reported by AEC (1949a) to estimate intakes 

during uranium operations. Other Linde particulate air concentration 

data were reviewed and considered, but are not discussed here. (p. 30) 

What are the other particulate air concentration data reviewed and considered by 

NIOSH/ORAU? Do these data have higher or lower values than the “largest time 

weighted air concentration?” 
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11. Section 3.3.1, Radon Breath Data, states: “Room background concentrations of radon 

were sometimes subtracted from the results and sometimes not” (p. 31, 2nd para.). How 

did NIOSH/ORAU handle these variations in data? 

12. Section 3.3.1, Radon Breath Data, states: 

Radon breath results are the starting point for dose reconstruction, so it 

might be necessary to back-calculate breath results from either Ra-226 

burdens or tolerance levels. The original records might have sufficient 

information for this determination. This information has not yet been 

tabulated for inclusion in the site profile. (p. 31, 2nd para.) 

Why might the original records have sufficient information for this determination? Will 

the information be included in future revisions of the Site Profile? 

13. Section 3.4, Uranium Progeny, states: 

This along with the claimant-favorable assumptions made in the estimation of 

worker dust exposures is judged to provide sufficient overestimation to balance 

any underestimation associated with the handling of waste products. (p. 31) 

On what basis does NIOSH/ORAU think the assumptions made in the worker dust 

exposures are claimant favorable? Are they not “time-average” values? Can 

NIOSH/ORAU list the areas of underestimation associated with the handling of waste 

products? 

14. How did NIOSH/ORAU arrive at the 22.4 pCi/L value (p. 33) as radon exposure for 

workers who did not work or have their offices in the process buildings prior to standby? 

15. Section 3.5.3, Tonawanda Laboratory Radon Exposures, states: 

For dose reconstruction, it is assumed that African ore processing at 

Tonawanda Laboratory occurred during its whole MED period, but that 

the peak concentrations of radon were equal to the medium values at the 

Ceramics Plant. It is assumed that after MED research and initial 

cleanup at the Laboratory ended, the radon concentration dropped to 

10 pCi/L, the level in the Ceramics Plant after it switched from African 

ore processing to domestic ore processing, and remained there until the 

end of the cleanup at the Laboratory. After the end of its cleanup, radon 

exposure in the Laboratory was based on the highest geometric mean 

PAEC determined for a Tonawanda site building from measurements 

made in 1981, 1.68E-02 WL for Building 31 (based on analysis of data in 

BNI 1982, Table B-3). .... Radon exposure rates provided here are annual 

PAEE rates in WLM/y. Each value is assumed to be the median of a 

lognormal distribution with a GSD of 3.43. This GSD is based on the 

location category having the highest GSD (3.43 per Table 6). (p. 34) 

Can NIOSH/ORAU explain what is meant by “the peak concentrations of radon were 

equal to the medium values?” What is the highest geometric mean PAEC (potential 
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alpha energy concentration)? What are PAEE rates? Why is the median of a lognormal 

distribution with a GSD of 3.43 claimant favorable for Linde dose reconstruction? 

16. Section 3.7, Ingestion Intake Estimates, states: 

NIOSH (2004) indicates that the ingestion rate, in terms of dpm for an 

8-hour workday, can be estimated by multiplying the air concentration in 

dpm per cubic meter by a factor of 0.2, so the uranium ingestion rate 

based on an air concentration of 7 alpha dpm/m3 would be 

0.563 dpm/workday. To adjust this to ingestion intake per calendar day, 

0.563 dpm/workday is multiplied by 250 workdays per year and divided by 

365 days per year, which equals 0.385 dpm/day. For the assumed 

exposure at 33 MAC, no alpha activity is apportioned to progeny so the 

daily uranium intake would be 316 dpm/day. (p. 35) 

What is the basis for the factor of 0.2 used for ingestion intake? Why is the assumption 

of an 8-hour workday, 250 workdays per year (i.e., 2,000 WH/y), seen as claimant 

favorable when it is likely that many workers put in more hours?

17. Section 3.8, Consideration of Bioassay Data, states: 

Given a chronic exposure to uranium and its alpha emitting progeny at 

300 MAC, the activity fraction of Ra-226 would be 0.196, which means 

that the chronic inhalation rate would be 2.7E+04 dpm/. This gives a 

whole body activity of 2.6E+05 dpm at one year, and about 4.0E+05 dpm 

at 4 years (calculated using IMBA Expert (OCAS), Version 3.2.20). The 

Ra-226 body activity was estimated using the largest breath radon result 

found for Linde, 2.2 pCi/L, by multiplying the radon result by a conversion 

factor of 2.52E+05 pCi/(pCi/L) (ORAUT 2005). This gives a body activity 

of 5.5 E+05 pCi, which is equal to 1.2 E+06 dpm, and is within a factor of 

3 of the estimated intake from a 4-year chronic exposure to 300 MAC.

Because other Linde radon breath analyses are lower and because a 

chronic exposure scenario may not best represent a worker’s exposure 

pattern, the assumption of 300 MAC chronic exposure was believed to be 

adequate for reconstructing doses in the pre-1947 research and 

production period, but at this time this period is reserved. (p. 36) 

Can NIOSH/ORAU provide the calculation sheets for further review, including all 

parameters used? Given the concern with chronic exposure scenario, did NIOSH/ORAU 

develop a list of episodic or acute exposure scenarios at Linde? 

18. In Section 3.9, Occupational Internal Dose Reconstruction Assumptions and Summary 

1942–1954: 

a. What is meant by “for workers, whose work location is considered indeterminate, 

intakes for the Ceramics Plant are assumed” (p. 36)? What are the assumed intakes? 

Why is this approach claimant favorable? What intakes should the dose reconstructor 
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use if a worker’s work location can be determined, since the Site Profile does not 

provide intake value by location? 

b. Why is the “dose distribution for particulate intakes ... assumed to be constant” (p. 

36). Is this approach claimant favorable? Is 3.43 the largest GSD value? How does 

it compare with the 95th percentile value? 

c. The retired workers and site experts SC&A interviewed indicated that they had never 

undergone any bioassay or urinalysis while they worked at Linde Site. Do the Linde 

employee files reflect this comment? 

d. Urinalyses for uranium until 1950 were suspect (see page 40 of the Mallinckrodt Site 

Profile). Do NIOSH and ORAU feel confident that the Linde data could be used to 

produce claimant-favorable dose reconstruction? 

Occupational External Doses 

1. The Section on Beta Radiation states that “Workers who frequently handled significant 

quantities of beta-emitting materials were assumed to have higher beta doses to the hands 

and forearms than to the remainder of the body (p. 38).” Did NIOSH/ORAU consider 

that these workers might have carried burlap bags containing radioactive material close to 

their chests when they were loading or unloading the uranium ores? The site experts 

SC&A interviewed indicated that workers frequently sat on the burlap bags during lunch 

and break times for many years, even after the MED project was completed. Did 

NIOSH/ORAU consider exposures to workers from these bags? 

2. Section 4.4, External Dose Reconstruction Summary 1942–1954, states: 

Each estimate is considered to be the median of a lognormal distribution. 

A GSD of 3.0 is assigned to all beta and gamma dose estimates based on 

the typical GSD levels estimated for the underlying data, i.e., beta and 

gamma doses are assigned as lognormal distributions with GSDs of 3, and 

an acute exposure rate. Neutron doses are assigned as a constant 

distribution and a chronic exposure rate (p. 63) 

How is this approach, assigning the median of a lognormal distribution as the bounding 

dose value, claimant favorable for determining beta, gamma, and neutron doses? Other 

examples include: 

a. For Step I process workers, NIOSH/ORAU assigns an exposure rate of 5.35 R/y to all 

workers for dose reconstruction purposes (based on Table 24). Again, this exposure 

rate for the loaders is the median of a lognormal distribution with a GSD of 2.61. 

b. For the rehabilitation and production period from 1947 to 1949, the Site Profile 

assumes the medium beta dose category is 1.95 R/y and the low beta dose category 

1.0 R/y (p. 57). 

c. For the rehabilitation and production period from 1947 to 1949, NIOSH/ORAU 

assumes the medium gamma dose category to be equal to the highest gamma dose 

rates of 1.61 R/y (p. 59). 
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3. Section 4.1.2.1, Pre-production, 1942–1943, presents external exposures during nonproduction periods at Linde. It states: “The beta dose rate was assumed to apply to the 

whole body with no added dose to the hands and forearms because it is unlikely that there 

was any significant handling of radioactive materials by Ceramics Plant workers in this 

period (p. 40).” Didn’t some workers transport and unload the uranium ores from 

boxcars in this period?

4. Can NIOSH and ORAU provide support to the external exposure time given in 

Section 4.1.2.1 (p. 40) of 0.5 h/d, which totals 150 hours per year, from 1942 to 1943? 

5. Section 4.1.3, Production, 1943–1946, estimates worker doses in subsection 4.1.3.1.1. 

The latter states: 

In order to use the data in Table 19 to estimate time-averaged beta dose 

rates applicable to the entire 1943–1946 production period at Linde, the 

ratio of the average dose rates to those measured in Table 19 was 

estimated. Table 20 documents the determination of the ratio. The grade 

range of the L-30 ore used at Linde was 8-12% (Aerospace Corporation 

1981, Table B-1). To obtain the highest ratio, it was assumed that the 

measurements in Table 19 were made on the lowest L-30 ore grade, 8%. 

It was also assumed that beta dose was proportional to electron energy 

released per decay and the worker doses were proportional to the mass of 

ore processed. With these assumptions, it was estimated that average 

doses would have been 0.84 of the doses predicted by using the data in 

Table 19. Therefore, the estimated dose rates in Table 19 were multiplied 

by 0.84 to obtain time-averaged dose rates for 1943–1946 production. 

The results are in Table 21, which summarizes the results of all beta dose 

rate estimates in this section and groups the job categories into three 

groups (high, medium, and low). (p. 45) 

Data in Table 19 (p. 44) represent beta exposure rates based on a 1-day radiation survey 

of the Linde facilities on January 23, 1944, during processing of a particular batch of 

African ore. How can NIOSH/ORAU justify using this single data point (and, how is it 

claimant favorable?) to estimate time-averaged beta dose rates for the entire 1943–1946 

production period at Linde? What correction factors did NIOSH and ORAU make to 

account for uncertainties? Please define these three groups of job categories associated 

with beta dose rates. 

6. The Estimation of Worker Doses section (Section 4.1.2.1.1) states: 

Workers in 1943–1946 who held jobs analogous to ones in 1947–1949 

were assigned three times the 1947–1949 beta dose rates. The factor three 

increase accounts for potential exposure to radiation from waste products 

from unrefined uranium ore and for the possibility that procedures in 

1943–1946 did not involve as much radiation protection. (p. 45) 

What is the justification for the factor of three; is it claimant favorable? 
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7. Table 33 (p. 60) of Section 4.2, Tonawanda Laboratory Beta and Gamma Exposure, 

assumes the 1942–1954 beta and extremity doses to workers to be one half of the value of 

the most exposed Ceramics Plant process workers during the 1943–1946 production 

period. What bases support this assumption as claimant favorable? 

8. What factors did NOSH/ORAU use for correcting film badge underestimates of lowenergy photon exposures at Linde? Is NIOSH/ORAU planning to publish a Technical 

Information Bulletin similar to ORAUT-OTIB-0010, 1/12/2004 Rev. 00, for Atomic 

Weapons facilities for the period prior to 1970? 

9. What type of film badges were used at Linde Site? Can sensitivity data be provided? 

Were TLDs used in the later years of the cleanup period? 

10. What types of survey equipment were used by MED/AEC/NYOO at Linde? What were 

the LODs for the equipment? 

Occupational Medical Doses 

No comments. 

Residual Contamination After 1954 

1. Section 6.1.1, Internal Exposure, provides Table 38 (p. 72) displaying the highest 

monthly average air concentration of Ra-226, Th-230, and U-238 for July 2000–June 

2004. How should dose reconstructors use these data in the period of 2000–2004 for 

determining internal doses for cleanup workers in 1950s, 1960s, 1970s, 1980s, and 1990s? 

2. Section 6.1.1, Internal Exposure, states: 

Building 30 was found in 1976 to be the most contaminated building on 

the site (ORNL 1978). The indoor airborne uranium concentration in 

1976 was measured as 1.90E-02 pCi U/m3

 (ORNL 1978, Table 14). This 

larger uranium air concentration is used to estimate a chronic intake. To 

maintain the consistent intake units and estimate the annual intake, the 

uranium air concentration of 1.90E-02 pCi/m3

 is multiplied by 

2.22 dpm/pCi, a breathing rate of 1.2 m3

/h and 2000 work-hours/y. The 

annual intake rate is estimated at 100 dpm/y for all Tonawanda employees 

beginning in 1955. Dividing by 365 days/y give a daily uranium 

inhalation intake rate of 0.277 dpm/day. The Th-230 and Ra-226 daily 

inhalation intake rates would be 0.233 and 0.471 dpm/d, respectively. 

Ingestion intake rates are estimated using the steps described in 

Section 3.7. The summarized intake results are shown in Table 39. It is 

unlikely that uranium would be in a chemical form consistent with type F 

absorption during the residual contamination period, so it is assumed that 

only types M and S would be inhaled after 1954. Th-230 could be type M 

or S, and Ra-226 is type M. (p. 72) 
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There were other radiation surveys done at Linde before 1976. Did NIOSH/ORAU 

compare this “larger uranium air concentration” (2nd sentence) to those survey data? Is 

there a higher concentration than 1.90E-02 pCi U/m3, since it is denoted as the larger, not 

the largest? The paragraph following the above states: 

Workers noted that Building 30 renovations occurred in the 1960s that 

could have influenced air concentrations. Specific details of the 

renovation, including the actual period of renovation, dust control 

measures, location of work and occupancy of areas are not available. It is 

reasonable to assume that renovations could have resulted in elevated 

airborne radioactivity. (p. 72) 

With these renovations in mind, how could the Site Profile assume that this “larger 

uranium air concentration” could bound the airborne concentration value? Did 

NIOSH/ORAU perform any uncertainty analysis of the concentration data? 

3. The text following the above indicates that a factor of 14 was assumed to be reasonable to 

describe the uncertainty at the 95th percentile associated with the possibility of elevated 

intakes during Building 30 renovations (p. 72). How did NIOSH/ORAU arrive at this 

factor of 14? 

4. What does the Site Profile mean by “other progeny are partially accounted for by the 

assigned radon exposure (p. 72)?” What are these progenies and how is the “partially” 

compensated for? 

5. Section 6.1.1 subsection, Radon Daughters, estimates the radon exposure post-1954 to be 

2.01E-01 WLM/y, based on 55 measurements of radon daughter concentration in 1976 

and 1981 (p. 73). This value is derived by multiplying 12 work-months per year by the 

geometric mean PAEC of 1.68E-02 WL, with a GSD of 1.89. Is this radon exposure 

claimant favorable or bounding for worker exposure in the 1950s, 1960s, and early 1970s? 

6. Section 6.1.2, External Beta and Gamma Exposure (post-1954), states: 

BNI 1982 (p. B-12) stated that the points of maximum radiation 

determined in the ground surveys were slightly displaced from but in 

general agreement with those determined in a 1979 aerial radiological 

survey. The total number of readings =25 μR/h reported by BNI was 16. 

The net readings (after subtraction of 8 μR/h to correct for background) 

had a geometric mean of 94.0 μR/h and a GSD of 3.95. This was taken as 

an estimate of worker exposure rate when outdoors. This estimate was 

assumed to apply from January 1, 1955 to the present (2005). (p. 73) 

Why is the geometric mean of 94.0 μR/h considered claimant favorable for estimating 

worker outdoors exposure rates, especially in the 1950s, 1960s, and 1970s? Is there any 

correction made for natural attenuation over that time period?

7. Table 39 in Section 6.3, Dose Reconstruction Summary, 1955 to Present, shows annual 

internal and external exposures to residual activity (p. 74). Can NIOSH/ORAU explain 
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how this table was developed? Where do the penetrating and non-penetrating exposure 

values come from? 

Other Issues 

1. Section 2.3.4, Other Radiological Activities, states: 

In 1949, Linde workers were used to unload drums of K-65 shipped by rail 

to the Lake Ontario Ordnance Works at Modeltown, New York (Wolf 1949, 

Heatherton 1949a). Linde film badges were worn during this work. (p. 22) 

How were the exposure received by these Linde workers at the LOOW handled in the 

Site Profile? Were the workers credited with the exposure? 

2. The above section states (p. 21) that radioactive liquid wastes were discharges into onsite 

injection wells. These wells sometimes overflowed and became clogged; workers were 

required to descend into these wells to perform maintenance and unclog them. Did 

NIOSH/ORAU consider the exposure of workers to radioactive waste inside the wells? 

3. The text following the above states: 

The estimates of uranium exposure based on estimates of exposure periods 

and source term, which were based on worst case assumptions when a 

parameter was not well supported by available information would be 

sufficiently bounding to account for small amounts of radioactive 

strontium. (p. 22) 

What are the worst-case assumptions referenced? How much is a small amount? 

4. Section 2.6, Decontamination During the MED/AEC Contract Period, states: “a 

systematic radiation survey was conducted to identify area of contamination” (p. 25). 

However, this systematic survey was not performed until the entire building was vacuum 

cleaned and flushed with water. For this reason, the true contamination level after the 

production process ended in the facilities is not known. How does NIOSH/ORAU 

estimate the potential internal and external doses to workers during the production period? 

How did NIOSH/ORAU estimate the potential internal and external doses to workers 

performing the vacuuming and flushing? How were the radioactive wastes handled 

during the vacuuming and flushing process?

5. The above section goes on to say:

The memorandum reported “overall” floor and wall levels to be 

1.01 “mreps/hr/ft2

” beta plus gamma. This was considered excessive. The 

memorandum cited data indicating that this could be reduced to 

0.065 (mreps/hr)/ft2 by covering contaminated floor areas with asphalt tile. (p. 25) 

How did NIOSH/ORAU use this information in the dose reconstruction for cleanup 

workers and also the workers later occupying Building 38? How should the dose 

reconstructor use this information for the cleanup workers? 
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6. Section 2.7, Post MED/AEC Operations (p. 26), discusses elevated employee radiation 

exposures and residual radioactive contamination associated with areas in or near 

Buildings 57, 58, and 90. What were these three buildings used for? They are not 

discussed or described in Section 2.1. What are the radioactive contaminants associated 

with these buildings? What types of worker radiation exposures were there? How high 

were these exposures? 

7. In the late 1940s, the Manhattan Engineering District (MED) established permissible 

doses for external radiation in places where workers were handling and processing 

uranium and its compounds. These limits were 500 mR beta equivalents per day or 3,500 

mR beta equivalents per week. How are these two dose limits used in the dose 

reconstruction for Linde? Did NIOSH/ORAU take these two dose limits into account 

when considering the bounding or worst-case scenarios? 
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ATTACHMENT 2: ADVANCE QUESTIONS FOR SITE EXPERT 

INTERVIEWS 

General Worker Questions 

1. Which plant did you work in and what tasks did you perform. Did these tasks involve 

radioactive material? If so, what type? 

2. When did the AEC work at Tonawanda Laboratory begin and end?

3. When did Ceramics Plant personnel start to handle radioactive material? 

4. How many people worked at Linde during the AEC years? 

5. Describe the layout of the facility (e.g., buildings, processes, etc.)? Describe the work with 

radioactive material in the following buildings. During what years was this work performed. 

• Building A 

• Building B 

• Building 14 (Tonawanda Laboratory) 

• Building 30 

• Building 31 

• Building 37 

• Building 38 

• Storage Areas 

• Waste Disposal Areas 

6. Did the Tonawanda Laboratory serve as a pilot plant for the development of uranium 

processing methods? 

7. Describe the process flow for each step of uranium processing, how long each process took, 

and what it involved? 

8. Where did you receive the uranium ore from? 

9. How was the uranium transported to and from the site? What type of containers used for 

transportation of the ores? 

10. Who were the railroad workers employed by? 

11. Who opened the boxcars when they first arrived? How was this done? 

12. Where were uranium ores stored? How were they stored? 

13. How were these uranium ores transported from the boxcars to the staging areas? To the 

storage areas? To the process areas? 
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14. How were uranium ores taken out of the containers? What equipment was used? Where were 

the workers involved? Were they wearing special protective equipment? 

15. Were there any security or entry logs maintained for any areas of the plant? 

16. Describe “high risk jobs” (i.e., jobs that posed a higher level of exposure risk), how many, 

and what type of workers were involved with these jobs? 

17. Were employees frequently transferred between the different processes or did they maintain 

the same job for a majority of their career? Did your tasks and assignments change routinely? 

18. Was there daily movement between the laboratory and the ceramics plant? 

19. How many hours a day and per week did you routinely work? Did your work include work 

on Saturdays, Sundays and/or holidays? 

20. Where was support personnel (e.g., health physics technicians, foreman, security, etc.) 

located in relation to the immediate production lines, machining or maintenance? 

21. What was the relationship between the administrative and production areas? Were these 

areas attached to one another? Did administrative folks enter the production area? 

22. Was production paperwork sent from the production areas into the administrative areas? 

Was this paperwork surveyed to determine if it was contaminated? 

23. Were there operations at Linde that involved enriched uranium (e.g., higher in U-235 

content)? 

24. Were there any operations at the plant involving thorium processing? 

25. How long was the uranium stored prior to and after processing at the facility? Where was it 

stored, and how much contact did workers have with stored material? 

26. Were there recycling operations at Linde? If so, please explain. 

27. Is the information provided in Table 2 of the site profile complete? 

28. Where was the waste stored and/or disposed of from the processes? 

29. Were you ever involved in waste clean-up operations and if so describe? 

30. Was there a tailings pond? 

31. Explain the process of abrading? 

32. Upon completion of operations what was done with the equipment? 
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33. Was there any regular facility sweeping and dusting done? If so, how often? When it was 

done, how was it done? Were the workers wearing any protective equipment? 

34. During the early (1940s/1950s) remediation, for those buildings not demolished, was the 

ventilation system removed, replaced or decontaminated? 

35. What is the status of Buildings 31 and 37? When were they remediated? 

36. What chemical exposures have you received in conjunction with radiation exposure? Were 

these chemical exposures simultaneous with radiation exposure? 

37. What was the corporate attitude towards safety of the worker? How did this change over 

time? 

38. Was there any worker training in the areas of personnel safety, operation, and equipment use? 

39. Was any protective equipment, such as glove, PC, respirator, dosimeter, safety glass, 

provided to the workers? 

External Exposure 

1. The early monitoring programs monitored only a portion of the workforce. Which workers 

were monitored? 

2. Did you ever wear pocket ionization chambers (also called PICs or pencils) at the plant? 

3. Were you aware of the use of co-worker exposure data to assign a dose? If so, what 

problems or advantages do you see in doing this? 

4. How were individuals directed to wear their beta/gamma dosimeter? 

5. Where were the dosimeters stored? Were they allowed to be taken home? 

6. Were workers ever given neutron badges? 

7. Where were badges stored historically? Were individuals allowed to take them home? 

8. Was extremity dosimetry assigned to workers? 

9. Are/were there conditions where partial body exposures (e.g., skin, lens of eye, gonads, chest, 

etc.) would have significantly exceeded whole body exposure measured by the routine 

dosimeter? If so, explain. 

10. Was/Is there an area dosimetry program at Linde? 

11. Was the time you spent in particular areas limited in any way? If so, under what conditions? 

How was the time spent in the area monitored? 
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12. Are you aware historically of unsanctioned practices by workers with respect to their 

dosimeter or bioassay samples (e.g., putting them on sources, not wearing them, etc.)? 

13. Specifically, did you ever sit on the uranium? Did other workers sit on the uranium? Was 

there a lot of lifting of uranium metal pieces and setting them down? 

14. What operations at the Linde plant are manual versus automated? What operations involved 

hands-on work with uranium or other radioactive material? 

15. How many hours in an 8-hour period did you have direct contact with uranium or associated 

waste? 

16. Were workers surveyed before they began the work day and surveyed before they left the 

plant? 

17. Was there any area survey done by the health physics technicians? If so, how often was it 

done? Daily? Weekly? Monthly? 

Radiation Operations 

1. Did you know you were working with radioactive material? 

2. Did you receive radiological worker training? If so, when did this training begin? 

3. Did you have a plant health physicist or safety officer? If so, who was this? 

4. Did Linde use Radiation Work Permits or Special Work Permits? If so, did these permits 

include requirements for dosimetry and minimizing dose? 

5. Did Linde use postings to alert workers when they were entering radiation areas? 

6. Is there field characterization data? If so, where is this data located? How can we obtain a 

copy? 

7. Was x-ray equipment or radioactive sources used to examine the uranium for quality 

purposes? 

8. Did Linde Plant handle radionuclides other than uranium? This would include Research and 

Development activities. If so, what were these radionuclides and in what quantity were they 

handled? 

9. Workers indicated to NIOSH that there was strontium at Linde. Can you please provide 

more details? 

10. When did Linde upgrade the ventilation system to support uranium work? If so, what type of 

ventilation system was used? Was there any filtering capability, like HEPA filter? 
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11. Was there any air sampling taken by the health physics technicians? If so, how often it was 

done? Where were the sampling results kept? 

12. What types of maintenance work involved with the uranium operation and process? Did 

these maintenance workers require the support of mechanics or electricians? How close were 

these workers to the processing uranium streams? 

13. How were liquid wastes handled in the operation? 

14. How were solid wastes handled in the operation? 

Internal Dosimetry 

1. Which jobs involved heavy work (e.g., strenuous labor)? 

2. Which jobs involved light work? 

3. Did you ever work with the lathes, grinders or milling machines? If so describe your work 

and what you perceived as potential hazards? 

4. Did you work in areas with visible dust? 

5. Were your lungs irritated by the dust you inhaled? 

6. Did you wear any respirator? If so, what type of respirator? 

7. What were the floors of the building made of? Were the floors cleaned regularly to reduce 

dust in the general areas? 

8. Where did you take your coffee or lunch breaks? 

9. Were there showers or change rooms at the facility? Did you shower and change your work 

clothing before leaving for home? 

10. What type of protective equipment did you wear? (e.g., respirators, hoods, painters masks, 

gloves, company issued clothes, hats, booties, safety shoes, goggles, face shields, aprons, lab 

coats, etc.) 

11. Were there situations where an individual in the immediate vicinity of the work wore 

protective equipment yet an individual standing near that individual did not? If so, please 

explain? 

12. How were respirators assigned? Did an individual use the same respirator more than one 

time? If so, where was the respirator stored in between uses? 

13. What particular forms of uranium did you work with and were there situations where you felt 

you were likely to receive a higher dose than normal? 
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14. Did you hear of or know of uranium fires? Do you remember anyone discussing uranium 

hydride or spontaneous combustion problems? 

15. Did you fire the uranium oxide? In what set was this done? At what temperature? 

16. Historically, who was put on a bioassay program (i.e. submitting urine samples for detection 

of radiation)? 

17. Were there any employees that were exposed to radionuclides who were not on an adequate 

bioassay program? 

18. For those monitored via urinalysis, what directions were given for submitting the samples? 

Were you asked to wait for a specific period before collecting the sample or could you start 

right away? 

19. Did safety or outside organizations perform routine air sampling of the general area and 

worker breathing zones? 

20. Are thoron and radon occupational exposure issues in any areas of the facility? If so, where? 

21. Were you exposed to fumes? If so, explain. 

22. Did Linde Plant process and/or store recycled uranium? 

23. Do you feel NIOSH integrated worker input into the site profile? 

24. Was there a clean room assigned for worker lunching and drinking? Where did you keep 

your lunch bags? 

25. Did you change into work clothes when you entered the plant? If not, how was the condition 

of your clothes when you arrived home? Dusty? 

Medical Exposure 

1. What were the medical exam requirements for workers in the present and historically? Did 

workers receive medical exams and if so what did these exams involve? 

2. What did the medical exams include? 

3. Did workers receive medical chest x-rays while at Linde and if so how often was that 

typically done? 

4. Where were chest x-rays done? 

5. How frequently did Linde employees have medical exams? 

6. How frequently were they given x-rays as a part of this exam? 
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7. How many x-rays were taken per exam? Which direction were you facing? 

8. Did you receive pelvic x-rays? 

9. Was photofluorography used at any time? Were 4” x 5” films ever used for x-rays? 

10. Are there other items which you would like to add that would help evaluate medical x-ray 

exposure? 

11. Have you seen your personal record file? Were you ever provided with your dose record? 

How often it was provided? 

12. Was there any fire hazard in the operation? Please explain. 

Environmental Monitoring 

1. How were these environmental releases documented? Where are these records located? 

2. Are you aware of historical particulate and gaseous releases from facilities to the 

environment? If so, describe these. 

3. Did Linde have a formal environmental monitoring program in place during the production 

years? During remediation years? 

4. What is the extent of outdoor contamination at Linde in the soil, groundwater, vegetation, 

etc.? (Both historically and currently) 

5. What is the average distance of the plant buildings from the site boundaries? How close is the 

site boundary from the neighboring residential areas? 

6. What is the relationship between Linde and the state of New York? Is the state actively 

involved in oversight of work at Linde? 

7. What EPA-regulated activities are currently occurring at Linde? 

8. Were there radiological releases from Linde facilities other than uranium? If so, what 

radionuclides were involved and when did these releases occur? 

9. Were there any significant episodic releases from Linde? 

10. What controls did Linde implement over time to reduce environmental emissions from the 

facility? Were these controls effective? 

11. Were there any burning operations at Linde involving radioactive material? If so, please 

explain. 
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Records 

1. Who currently maintains the historical records related to Linde operations? 

2. Do radiation exposure records exist? If so, what type of records (e.g., bioassay data dosimeter 

records, air sampling, radon breath analysis, radiation surveys, etc.)? How complete are 

these records? 

3. Were your dosimeter and bioassay results made available to you routinely? When did this 

begin? 

4. Are you aware of any code names encountered in the records and their meanings? 

5. How were incident, spills, personnel contamination, etc. documented? Where are these 

reports stored? 

6. Has dose of record been modified for any reason over time? 

7. Has the site destroyed any dosimetry, bioassay, or field radiological records in the past? If so, 

what were they? 

8. Can you recommend resources (i.e., technical reports, books, films, etc.) that may be helpful 

in understanding the historical operations of the site? 

9. When you left the facility, did they provide you a copy of your personal dosimetry record? 

Facility Clean Up Operation 

1. What was done during cleanup as compared to D & D? 

2. Were facility sweeping and dusting and vacuuming done? 

3. How were the wastes generated handled? 

4. Were workers provided with respirator and other protective equipment? 

Facility Decontamination and Decommissioning Operation 

1. Was sandblasting used in the D&D of the facility? If so, where was it used? 

2. Was there high dust level during the sandblasting process? Was water mist used to reduce 

dust level? 

3. Were respirator or protective equipment provided to workers during sandblasting? 

4. What else was done during D &D? 

5. Was there any air monitoring, personnel survey, and area survey done during D & D? 
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ATTACHMENT 3: SUMMARY OF SITE EXPERT INTERVIEWS 

FORMER WORKERS 

Interviews were conducted with seventeen former Linde Ceramics Plant and Tonawanda 

Laboratory employees. Years worked onsite at Linde Site represented by those interviewed 

range from 1951 to 1998. The interviews were conducted by Desmond Chan and Kathryn 

Robertson-DeMers, members of the SC&A Linde review team. The purpose of these interviews 

was to receive first-hand accounts of safety and personnel monitoring practices at Linde and to 

better understand how operations and remediation activities were conducted. Interviews were 

done in person on January 12–14, 2006, in Buffalo, New York. Interviewees were selected to 

represent a reasonable cross-section of production areas and job categories. Interviewees were 

originally obtained through the FACTS (“For A Clean Tonawanda Site”) organization and the 

NIOSH/ORAU team. An effort was made to contact living safety staff; however, SC&A was 

unsuccessful in locating them. A brief tour of the perimeter of the Linde Site was conducted to 

visualize the proximity of various operations, the level of remediation underway, and the 

structure of buildings still existing from the Manhattan Engineering District (MED) era. 

Although SC&A tried to contact personnel from Linde who worked on the MED, we were 

unsuccessful in finding individuals who were alive and coherent by the completion of this 

summary. 

Workers were briefed on the purpose of the interviews and the Linde Site Profile. They were 

asked to provide their names in case there were follow-up questions. Participants were reminded 

that they would be provided the opportunity to review the interview summaries prior to inclusion 

into this report. Interviewees were told that there were aspects of the Linde operations that were 

classified and this information could not be divulged. 

Former Linde workers interviewed worked throughout the Linde Site due to the nature of their 

jobs. Some of the primary buildings associated with their work included Buildings 1, 2, 2B, 8, 

10, 11, 14, 19, 27, 30, 31, 33, 34, 37, 38, 39, 43, 44A, 44B, 57, 59, 69, 70, 70A, 70B, 90, 98, 99, 

100, 100S, and 100N. There were only a few select operations on the site that were not readily 

accessible to all workers. The job categories represented in the interviews included: 

• Fabrication Shop and Maintenance 

• Tool and Die Making 

• Inspection 

• Laboratory Mechanic 

• Maintenance Foreman/Supervisor 

• Pipefitter 

• Draftsman/Designer 

• Building and Grounds Supervisor 

• Engineering Administration 

• Chemical Operator 
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• Heating, Ventilation and Air Conditioning Mechanic 

• Administrative Services (Office Manager, Secretary, Timekeeper, Accounting, Printing 

Department) 

• Research and Development 

• Project Management 

• Painter 

• Shear Operator 

• Truck Driver 

• Welder 

• Graphics Art Supervisor 

• Electrician 

• Millwright. 

The information the workers provided to SC&A has been invaluable in providing us with a 

working knowledge of the various site operations and the safety program. All interviews have 

been documented and summarized below. This summary is not a verbatim discussion, but is a 

summary of information from multiple interviews with many individuals. The information 

provided by the interviewees was based entirely on their personal experience at Linde Site. It is 

recognized that these site expert and former Linde worker recollections and statements may need 

to be further substantiated. However, they stand as critical operational feedback and reality 

reference checks. These interview summaries are provided in that context. Linde site expert 

input is similarly reflected in our discussion. With the preceding qualifications in mind, this 

summary has contributed to our findings and observations. 

General 

The MED project activities at the Linde Site involved the use of Buildings A, B, 14, 30, 31, 38, 

and 57. Buildings 31 and 38 were used for processing and production of uranium. Upstairs in 

Building 31, there was experimental work related to the project. The floors of most buildings 

were made of pitted and/or finished concrete. There were some areas with dirt floors. Buildings 

30 and 57 were set up to receive the uranium ore. The railroad tracks came right into the site for 

uranium ore deliveries. The railroad workers were not employed by Union Carbide. When the 

trains reached the gate, a Linde Train Operator would move the train from the gate into the 

Manufacturing Areas. Domestic ore was shipped in open boxcars. African ore was shipped in 

burlap bags which had to be unloaded manually. The end products were put into boxes which 

were stacked on railcars for shipping offsite to other facilities. 

Linde is in close proximity to the neighboring residential population and commercial structures. 

The site is located in the town of Tonawanda, New York. Building 30 was approximately 300–

400 feet from the site perimeter. Nuclear piles were put right next to the fence line. Rattlesnake 

Creek and Two Mile Creek both run through the length of Linde property. 
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The Linde Site is still operational although current operations do not involve radioactive material. 

The site has undergone and continues to undergo extensive remediation. There are significant 

levels of radioactive contamination in the parking lots, soil, wells, creeks passing through the 

property, and in some of the older buildings. Linde bought the golf course neighboring the 

facility due to environmental contamination. Part of the remediation of the site includes 

removing contamination from the creeks. 

Those accessing the Linde Site were required to have a company badge. Both salaried and 

hourly employees had access to nearly all buildings onsite. Their work took them to areas 

throughout the site requiring that they entered all buildings and production areas except that 

indicated below. Support staff such as engineering workers frequently visited the work areas and 

shops to ensure equipment and systems were working smoothly. Generally, the administrative 

support staff did not go into the production areas, but there were exceptions such as those that 

delivered mail or distributed paychecks. They delivered mails to all the onsite buildings, 

including offices, production areas, laboratories, receiving areas, shipping areas, and 

maintenance shops, at least twice every day. These mail delivery personnel were not wearing 

any special protective clothing or equipment. Very likely, they carried contaminants from 

production areas to offices and other areas within the Linde Site. 

It is important to note that Linde maintained a license to handle radioactive material after the end 

of the MED Project. This license was primarily for work being conducted onsite by the military. 

It is unknown what specific radionuclides were included in the military programs. Access to this 

building required a security clearance and the building was guarded. Most Linde employees did 

not have access to this building. The project involved testing special gas mixtures for deep sea 

divers. Krypton-85 gas was believed to be used in some military projects. 

There were also offices for administrative staff located in various production buildings, such as 

Building 14 and Building 30. A special engineering barrier or separation from production or 

operation areas was used. Office staff and operational workers were allowed to move freely 

between offices and production areas without egress monitoring or restriction. 

Site experts indicated that they worked the standard 40 hours per week plus 10%–20% average 

overtime. The amount of overtime varied based on the production schedule. Regularly, workers 

would work 8 hours per day with 45 minutes lunch break. In 1952–1954, there were workers 

working 7 days and 24 hours in Building 14 performing maintenance and chemical operations. 

Building 14 was not renovated before the new equipment was installed to produce silicon 

products. The old MED project system structures, including ventilators and piping, were still in 

place without undergoing cleanup or remediation. They used four crews with scheduled shifts 

for the entire operation in Building 14. Support workers were housed in the building at this time. 

In Building 52, there was a locker room in the northwest corner of the building. Workers would 

come through the guard shack (Building 23) to reach Building 52. They changed into work 

clothes in the locker room and then entered the production area to work. 

There were about 3,000 employees at Linde Site in the early 1950s. The number of employees 

gradually was reduced to 2,000 employees in the 1990s. 
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Site Description 

The southwest corner of the site was referred to as the Tonawanda Laboratory. This included 

Buildings 10, 11, 12, 13, 14, 15, 16, 23, 34, 43, 44, 47, 52, 62, 66 and 100. The factory area 

made up the remainder of the site and included the production facilities used during the MED 

project. Table A3.1 below describes the buildings on the Linde site and their current status. 

Table A3.1. Descriptions of Select Linde Buildings 

Building Description Current Status 

A Manhattan Engineering District offices Torn down 

B Administrative Torn down 

1 Administrative Support Current 

2 Factory Manufacturing Area Converted to laboratories; 

Current 

8 Old Powerhouse Pending tear down 

9 Meter and Generating Room Renovations 

10 Laboratory Administration with Labs; Navy 

Laboratory 

Current 

11 Instrument Service Department, Silicon Rubber, 

Silicon, Dynamonitor Lab, Navy Project 

Remodeled half the 

building in the late-1960s 

to create office area and 

large computer room; 

Remodeled again to 

remove the computer room 

and add labs upstairs 

14 R&D Laboratory, Uranium Processing; High Flux 

Tubing; Silicon Production; Experimental 

Laboratory; Oxygen manufacturing 

In Process of being Torn 

Down 

15 Engineering Laboratory; Cryogenic Laboratory; 

Garages; Krypton-85 gases used 

Renovations several times; 

Current 

16 Engineering Laboratory Renovations several times; 

Current 

19 Metal Fabrication Building Torn down; New 

Powerhouse built where 

this building was 

previously 

23 Guard Shack Torn down 

25 Battery House Torn down 

27 Engineering Department Current 

29 Maintenance Department Torn down 

30 Ceramics Building; Storage; Offices; wooden 

building with cinder blocks and dirt floor 

Torn Down 

31 Molecular Sieves; Research Laboratory; Uprocessing Pending tear down; 

Renovated in 1957 

33 Engineering Department Torn Down 

34 Molecular Sieves Laboratory Building Pending tear down 

37 Storage; High Flux Cooling Area Torn Down 

38 Rare Gas Laboratory; U-processing; wooden 

building with cinder blocks and dirt floor 

Torn Down 

39 Laboratory Building; wooden building with cinder 

blocks and dirt floor 

Torn Down 
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Table A3.1. Descriptions of Select Linde Buildings 

Building Description Current Status 

43/43A Machine Shop; Silicon Shipping Department; Waste 

Water 

Current 

43B/44 Animal Research; Navy OTEC Laboratory; 

Cryogenics Laboratory 

Current 

50 Car Shop Torn Down 

52 Gas Laboratory Current (see Bldg. 34) 

54 Furnace for Burning Chemicals (old incinerator) Torn Down 

57 Storage Area for finished products Torn Down; Connected to 

Bldg. 73 

58 High Pressure Testing Laboratory; bomb test Torn Down 

59 Nuclear Testing Laboratory; Source Storage Area; 

Cobalt Laboratory 

Torn Down 

60 Engineering Building Current 

62 Laboratory Fabrication Shop; Instrumentation and 

Works Engineering 

Laboratories were 

remodeled; Current 

67 Laboratory; Shelf Generating Reactor/Incinerator 

used for burning 

Renovated in the late1970s; Torn Down 

69 Tunnel Complex A portion of the tunnel has 

been removed. 

70 Truck Manufacturing; Acid Deck; UOP Testing Area Paved over the Acid Deck; 

Current; Renovated many 

times 

73/73A/73B Storage; Molecular Sieve Storage Area, Union Oil 

Products (UOP) operations; High-flux Operations; 

Offices – Three 

Renovations to connect to 

Building 57; Current 

74 Substation Current 

90 Warehouse Building Current 

Following the MED years, Linde was devoted to other projects. Laboratory personnel were 

involved in Research and Development on silicon rubber. They were experimenting with 

different materials to strengthen silicon. Under the UNOX Waste Water Treatment Projects 

project, they performed various tests on waste water to develop better ways of cleaning the water. 

The Union Oil Products (UOP) (later it was referred to as merely UOP) project coated tubing for 

the chemical industry. Building 54 housed a furnace which was used to burn chemicals and 

waste. 

Lathes, grinders, and milling machines were found in shops throughout the site. This type of 

evasive work occurred in the Machine and Instrument Shops located in Buildings 2, 11, 18, 19, 

37, 43, and 62. There was a Tool Crib in both Buildings 1 and 2, where machining and grinding 

of metal cutting tools (such as bits and drills) and equipment parts were done. 

Linde had a complete system of tunnels beneath the facility. These tunnels were referred to as 

Building 69. They were generally six feet in diameter and ran throughout the property 

connecting various buildings. The steam lines, piping, and other utilities were located inside the 

tunnels. These tunnels also provided walk and crawl spaces for maintenance workers. The main 

tunnel linked the Powerhouse (Building 8) to Building 39. The tunnel branches were connected 
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to other major buildings, including Buildings 2A, 8, 8A, 10, 11, 14, 30, 31, 38, 39, 58, 59, 70, 

and 75. 

Building 14 was a Research and Development Laboratory. It was involved in the purification of 

uranium ore during the MED era. The building had an open high bay, offices, and laboratories. 

There were change areas and showers within the building. The building had a large ventilator in 

the high bay without any special filtration system. The building housed a furnace and an arc 

furnace. After the MED project was completed, they decontaminated the building once. Later, 

they removed all the uranium production systems and laboratory equipment. They removed the 

concrete floor of the building due to contamination. But they did not remove the entire floor and 

also the contaminated walls. In addition, they did not remove the sanitary, sewage, storm water, 

and plumbing piping. The sanitary, sewage, and storm water lines went directly out to the public 

sewage system. The remaining piping was used and connected to new production systems in the 

building in later years. Although they went through this effort, a few years ago they found that 

the building was still contaminated. 

Linde was the world’s largest manufacturer and production plant for T-200 in the 1970s. They 

took part of the wall down between the two bays. The sheet metal was brought into Building 70 

from Yard 70 and put on the flame bed. The flame bed operator would square the sheet and then 

make or cut holes necessary to put piping and other material through. The roller was down in the 

pit. The sheet would be set on a ramp and pushed into the roller to make a shell casing. It was 

tack welded. It was then transported to Building 14 where it was welded with a big Boom 

Welder (suspended welding torch). The shell would not fit under the Boom Welder so a base 

was added to the Boom Welder to make it taller. Tracks were laid down for a trolley. The 

trolley transported the shell to the Boom Welder. The installation of the tracks required bolting 

the tracks to the floor. The offset was put in a pit. A shell (radius of 16’) was laid out and put on 

rollers. After this was completed, an offset roller could be set up. Holes were drilled in the floor 

with drills. The off setter was lowered into the pit and mounted to bolts in the wall. The shell 

was set in the off setter and the rollers were used to align the shells. The process was repeated 

several times to create a column. The final column when finished had a head welded on it. The 

shell was taken to the high or low bay shop in Building 2/2A. The welded casting was 

transported to the shop for partial assembly. The shell was then sent to Wickwire where 

additional parts were assembled. This required hands-on assembly. Final assembly was 

completed at Burn Harbor. No respiratory protection was worn during this operation. 

Building 14 was used for fabrication also. Bechtel had two experimental columns to be used for 

radioactive work. Part of the column was put down in the ground. Construction of several 

columns required specialized welders, fabricators, and metalsmiths. The construction effort 

involved several shifts of 4- or 5-man crews. 

In the late 1950s, the Stores Department was located in Building 30. This department shipped, 

received, and delivered parts, oxygen cylinders, and production units. Truck drivers delivered 

both clean and contaminated parts to all facilities at Linde. The clean and contaminated parts 

were all together in the truck during transportation. There was no specific precaution or 

instruction to handle them differently. 
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Early Uranium Production Years 

Each employee was provided one new set of work clothes every year, including a shirt, pants, 

goggles, leather gloves, and work shoes. There were no masks, laboratory coats, or shoe covers 

assigned to workers. They were not provided with any radiation film badge or monitor unless 

working on special projects and/or radiography. After working at Linde for more than one year, 

workers were provided with more than one set of work clothes. They would take one set home to 

wash and use the second set at work alternatively. When workers entered the buildings, they 

first went into the locker room and changed out of their street clothes into work clothes. 

During the MED project era, people going into and leaving the production areas in the Building 

31 were required to go through a monitor. The exact type of monitor is unknown but individuals 

had to clear the monitor prior to exit from this area. If they were not able to clear the monitor 

they had to shower again and pass through the monitor a second time. Shower facilities were 

available for about 100 workers to wash up at the end of the day before they left the Linde Site. 

Normally, workers did not want to wait in line to take shower. Many workers went home with 

their soiled and contaminated street clothes. There were no health physics personnel in the 

locker room to survey or check out the workers before they left. 

There were radiation signs posted during the production years in the MED project buildings, 

such as Buildings 14, 30, 31, 38, and 39. The postings were not specific about the type of 

radiation hazards involved. After 1952, many radiation signs were removed. The only radiation 

sign posted after 1952 was at the time of x-ray radiography work. All workers were told that 

there was no radiation hazard at Linde after the MED project was completed. There was no 

fixed radiation or airborne monitoring equipment used in any of the buildings.

During the early 1950s, the dust levels in Building 31 were so high that Linde brought in Dr. 

Joseph Shister from the University of Buffalo to evaluate the work areas. He confirmed that the 

dust levels were high and not safe for inside workers. 

In Building 30, they had the same dust problems. Dust would come off the overhead rafters. 

Sometimes, they would fall right onto workers’ lunch. During the MED period, they stacked all 

the contaminated burlap bags in storage area of Building 30. These contaminated burlap bags 

were kept in there until they were removed to be burned in the incinerator in the late 1950s. 

Many people working in Building 30, including operation personnel, secretaries, and 

maintenance workers, would sit on those bags resting or eating their lunch. This went on for 

many years. 

In order to lower the worker exposure during this production period, Linde did rotate people into 

the production work areas on a weekly and monthly basis. 

The union went through discussion and arbitration with Linde to improve the situation and to get 

more money for working in such a dirty and dusty environment. 

Office and other support staff were housed in Building 30. There were offices also in the 

second floor. In the south end of Building 30, they had a truck shop in 1960s. It had a room 
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with sealed, reinforced concrete retention walls for shielding purpose. This room was used to 

store African ore during the MED period. There were about 75–100 workers working in 

Building 30 at any time. 

Characterization 

Building 14 (Tonawanda Laboratory), Building 31, Main Building 30, Powerhouse, Building A 

(fabrication), Buildings 1 and 2 (fabrication), Building 37 (maintenance), Building 38, the soil 

staging area and parking lot outside Building 30, and the wells are areas of potential 

contamination through the years. Residual radiation was found in several buildings including 

Buildings 8, 14, 19, 30, 31 37, 38, 39, 57 (localized), and 59. 

In the back of Building 30, there were two areas where uranium ore was stored in about 1951. 

By 1954, this area had been cleared of the uranium. The ore was stored in canvas bags with 

about 50 lbs of ore each. There were several pallets of these bags. Individuals worked in the 

immediate area, some even having desks there. It was a very dusty location with only a door for 

ventilation. Individuals spent approximately half their time in this area. Site experts indicated 

that there were some radioactive rods [not fuel rods] stored in Building 30 in the corridor for 

some period of time. Individuals use to walk over these rods.

In surveys conducted by Oak Ridge and other organizations, Building 30 was found to be the 

most contaminated building onsite with respect to residual radiation. Building 30 also had a 

contaminated parking lot adjacent to it. It took six to seven years to remove the equipment from 

this building and renovate it. 

Radon exposure was clearly an issue at Linde. The level of Radon was dependent on the 

location onsite and the ventilation system in place during the years in question. Employees 

became concerned about the Radon levels at the site in the later years. One employee brought in 

a Radon detector from home. When the company found out they asked her what she was doing. 

They became very upset. 

115 core samples were taken throughout the site in the late 1980s. There was also drilling 

around the perimeter of the property. Samples have also been taken from storm sewers and run 

off water. In later years, the site safety staff was observed outside of 100 Building in white suits 

conducting sampling activities. 

The extent of the soil and well remediation during the initial cleanup period in 1949–1954 is 

unclear. In the 1970s, the Army Corp of Engineers did a baseline study of the contamination at 

Linde Site prior to cleanup of the facilities. The tunnel system at the site was initially believed to 

be uncontaminated. Later, contamination was found in this area. The Army Corp of Engineers 

continues to find contamination as they proceed with remediation. 

The former workers interviewed indicated that there was no plutonium ever used onsite at Linde 

Site. 
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Building Remediation and Remodeling 

There are a number of buildings on the Linde Site that have undergone renovation after the 

Manhattan Engineering District project to remove contamination and upgrade facilities. These 

renovations took place after the initial post-MED cleanup in the late 1940s/early 1950s. Site 

experts were involved in renovations on Buildings 9, 14, 15, 16, 30, 39, 57, 58, and 73. Several 

buildings including Building 14 and Building 30 were eventually torn down due to residual 

contamination. 

In 1957, Building 30 was renovated. Workers would be working in the building offices and in 

the receiving department on the first floor while the renovation workers were jack-hammering 

the contaminated concrete floor. Dust was kicked up everywhere. They removed the machines 

in the building and put new equipment in. In the process, dust came off from overhead beams 

and rafters. There was no cleanup of the dust cumulated in the overhead beams, thrust, shears, 

ceilings, walls, rafters, and structures when they completed the MED project. 

Building 30 underwent a second renovation from 1961–1968. During renovation, there were 

jack-hammering of the concrete/dirt floors, removal and replacement of beams, shearing or 

cutting of steel and aluminum materials, and spray painting over previously existing paint. 

Equipment such as cut off machines, shears, and saws were moved around. The primary concern 

with Building 30, which processed uranium during production years, during this period was the 

residual contamination both in and outside the building. There were several tanks underneath 

Building 30 that served as disposal tanks during the MED era. Shearing operations caused dust 

to drop down from the ceiling. During some remediation operations, such as removal of the 

rafters, safety set up pumps to monitor airborne dust. Building 30 had four Gravity Ventilators 

which were found to be contaminated. These ventilators worked similar to a chimney with very 

little particulate filtration. They were constantly sucking out the dust inside the building during 

the production years and during the remediation period. Concerns were expressed regarding 

roof-top contamination, environmental contamination, soil contamination, parking lot 

contamination, and offsite contamination. Sandblasting was used during the renovation of this 

building to remove contamination from the floors, walls, etc. Equipment was removed, or 

sanded and repainted to make it look like new. All radioactive material labels were removed at 

this time. Although the equipment and building was painted to contain the radioactive material, 

this process did not contain all the radioactive material. 

Contractors were hired to assist the Linde staff during renovations and remediation. Linde 

personnel provided on-the-job oversight of contractor activities. They were involved in handdigging dirt, cutting holes in roof and tiles, replacing roofing and windows, and other 

construction activities. Both Building 2 and 2A had their roofs removed and replaced because 

they were contaminated. 

Later on, Linde wanted to make the entire first floor offices with false ceilings. They ended up 

digging up the entire floor and the surrounding walls. The renovation process created so much 

dust for workers working in the building at the time. 

For Buildings 18, 19, and 57, there were five ventilators without special filtration system on the 

roof tops. They used cranes to take them down. After they took them down onto the ground, 
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maintenance used wired brushes to clean them inside out. The maintenance workers were not 

provided with any special protective equipment, face masks, gloves, clothes, shoes, or radiation 

monitors while they were performing the cleaning work. They were all contaminated. 

Facility Decontamination and Decommissioning Operation 

Residual contamination from the MED production years was and is still a significant issue at 

Linde. Site experts indicate that the entire Linde site was contaminated even following the initial 

cleanup of the site in the late 1940s and early 1950s. Contamination was not limited to the 

surface, but penetrated the ground. 

Waste generated as a part of the MED activities was moved several times to other areas onsite. 

Some of the contaminated wastes were dumped into the old quarry, which was opened in the 

early 1940s. This included waste material from production and contaminated dirt from 

remediation. There was also a yard storage area in this part of the plant. 

Prior to the construction of Building 90, the surface soil was removed due to contamination and 

moved near the fence line outside of Building 73. The material was manually shoveled and 

transported to this area. There was a main door that led out between the building and the fence. 

On summer days it was opened to cool the building. Food was grilled in this area. 

Eventually, the dirt was moved to an area near Buildings 75 and 76. They then moved this dirt 

to the Northwest corner of the property, and eventually offsite. Contaminated dirt from under 

Pad 19 was moved to the back of the property. The dirt piles were capped with dirt, covered 

with canvas, and landscaped with grass. Some of the waste shipped offsite was sent to 

Envirocare in Clive, Utah. Other waste was shipped to the Ashland sites and Lake Ontario 

Ordnance Works. The site experts referred to the onsite dirt piles as the “nuclear piles.” 

Material shipped from the site had a waste designation of 11E2. 11E2 waste is the waste 

produced by extraction or concentration of uranium or thorium from any ore processed primarily 

for its source materials (i.e., uranium, thorium or both). 

There were several waste disposal wells onsite. Two were located south of Building 8, one was 

near Building 30, one was near Building 39, one was in the vicinity of the Yard Storage, and one 

was at the side of Building 19 between the building and the railroad tracks. The well near 

Building 19 was filled with concrete. Uranium waste from processing was dumped down these 

wells. There was also a quarry located outside Building 31. There are no tailing ponds onsite. 

There is a buried vault on the property with unknown contents. A radiography company was 

brought in to locate the vault so it could be removed. They were not successful. 

Radiography and Sources 

There was an established radiography program for monitoring the quality of manufactured units 

and performing Research and Development. Standard x-ray units were also used for quality 

control. There were two sizes of film used for industrial radiography. The larger film was 

approximately 20” x 22”. The smaller film was lead packed and used to check seals. Film 

approximately 5” x 10” was procured in 1944. The purpose is unknown. The site housed a 
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sealed 90Sr source in Building 59 in a lead vault. Union Carbide used sources in the 1970s to 

study the effects of high radiation exposure on metal strength. They wanted to know how 

exposures affected the welds of material. The inspection department located in Building 31 

handled a portable 60Co. They used this 60Co source to inspect the wells, storage tanks, and 

trucking tanks. Engineering controls were in place to prevent excessive exposure from sources. 

Areas were roped off and signs were posted. 

Former radiographers indicated that there were hot spots under the vending machine and near the 

bathroom. As an experiment, one technician laid film under the vending machines with three 

lead pennies on top of the film. The film was left in place for two days. When the film was 

developed, there was an image where the penny had been placed. 

Radiation Safety 

The corporate philosophy towards industrial safety was stringent. Workers were required to 

wear safety glasses, safety shoes, and in some cases hardhats. There was fall protection 

equipment available if needed. Eye wash stations, guard rails, and toe plates were built into the 

facilities. Workers were told not to worry about radiation as the radioactivity was within the 

limits. No details were shared with the workers. 

Linde provided the workers with goggles, hardhats, gloves, welding masks, painter/dust masks, 

and coveralls. A majority of the workers wore clothes from home; although, some individuals 

were provided with coveralls or smocks. A locker room was available for individuals to change 

into work clothes and store their lunch and personal items. Showers were available for workers 

but they were not required to take a shower before they left the site. It often took more than one 

shower to get the powder out of your pores. Personnel monitoring (egress monitoring) was not 

required prior to leaving the site. There was no portal monitor used to screen workers entering 

and exiting work areas after the MED project was completed. There were times when outside 

contractors wore protective clothing and Linde employees did not. 

During work clothes would get extremely soiled. Workers washed their soiled work clothes at 

home. There was no instruction to separate work clothes from clothes of family members. 

There were various types of respiratory protection used ranging from dust masks to full-face 

respirators with cartridges. Although dust masks were disposable, they were used more than 

once by some workers. 

Although some stored lunches in their lockers, most workers ate their lunch at their own work 

site in the production areas or in the locker rooms. They seldom washed their hands before they 

ate. Coffee was available for workers. There were facilities where they could warm up soup. In 

the later years vending machines were installed. The food vendor for the company stored food in 

Building 37. 

Visible dust was noted during some jobs. When shearing took place in Building 30, dust would 

drop from the support beams above. Buildings 14 and 30 were particularly dirty. The dust 

irritated some individuals’ respiratory tract, especially where fine metal powders were located. 

Some site experts reported returning home having to blow dirt out of their nose to clear it. This 
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was especially true in Building 30 when shearing activities were underway. Dust would come 

down from the ceiling. Floors had to be swept two to three times per day to control the dust in 

some areas. 

Safety personnel were involved in periodic surveys and job-specific general area air sampling 

during the renovation period. For example, when they cut and replaced the rafters, safety ran a 

couple of air samplers in the vicinity of the work. These air samples were not within the 

breathing zone of the workers. In the early 1970s, a nuclear physicist was working at the Linde 

site. This individual controlled work related to pumps, tanks, and digging on the site. 

Periodically, he used a Geiger counter to evaluate whether work involved residual radioactivity. 

Although radiation was detected in some circumstances, the levels were not reported to the union 

or workers on the job of concern. 

There was not routine radiation monitoring during the breach of system in building previously 

used in the MED years. Maintenance workers removed piping and entire duct systems. The 

ducts were filthy and required vacuuming once per year. Standard filters and a limited number 

of dust bags were found in some facilities. Early ventilation was obtained primarily through 

windows and doors. 

Although site experts worked throughout the site, they did not recollect seeing radiation postings 

until the remediation process began. It is believed that the Army Corp of Engineers was 

responsible for their placement. At this point, some site experts recollected seeing postings in 

Buildings 14, 31, 10, 52, and 59. 

Note that during the tour of the site perimeter radiation postings were seen on the fence in certain 

areas. Workers were dressed in white coveralls, booties, and gloves. There was soil remediation 

in the area and an individual was monitoring the area with a Geiger-Mueller Counter. 

During the time period after the MED work, site experts indicated that Radiation Work Permits 

or Special Work Permits were not used. There was a Hazardous Work Permit for certain jobs 

which primarily dealt with potential industrial hazards. 

Personnel Monitoring 

External monitoring was limited to the X-ray Department and those individuals working in 

Building 31. Badge exchanges were as frequent as daily. Those who were assigned dosimeters 

stored them in a badge rack at the facility. They were instructed to wear the dosimeter between 

the neck and waist on the outside of clothing. Those site experts were not provided with 

radiation exposure reports, but were merely told their dose was within the limits. There was at 

least one situation in the early 1990s were a few workers received dose in excess of the limit. 

The cause of the higher dose was believed to be radiography related. 

In the 1950s, six (6) workers from Linde Site were sent to Lake Ontario Ordnance Work Plant to 

repackage K-65 to be shipped to Fernald. They were given film badges to wear during their 

work. No dose record was ever provided to these six workers after they finished this work. 
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Site experts indicated that pocket ionization chambers (i.e., pencils), extremity dosimetry, and 

timekeeping were not a part of the radiation safety program at Linde post-MED period. They did 

not recollect an area dosimetry program. 

None of the site experts interviewed submitted bioassay samples. There was monitoring during 

the MED work for personnel in Building 31. The facility also had a policy where overexposed 

individuals would be reassigned. 

Job-related safety meetings were held periodically, but often focused on industrial safety. When 

the union representatives asked about exposure to radiation, they were told by the safety 

representatives that radiation was not a problem. 

Environmental Monitoring 

There was no environmental monitoring system found within the boundaries of Linde Site. 

There was also no fence-line monitoring equipment used at Linde Site. In the later years, AEC 

and Army Corps of Engineers performed environmental surveys of the Linde properties and 

found radioactive contamination in creeks, rooftops, open areas, parking lots, soils, wells, and 

neighboring areas. 

There were four drainage sumps within the Linde Site. They were either at or near Buildings 

2/2A, 8/8A, 10, and 30. These sumps drained directly out to the Tonawanda City public sewage 

lines. There was no filtration or treatment systems built in these sumps. In addition, there were 

four injection wells within Linde for liquid waste disposal. These wells were at or around 

Buildings 8, 30, 73, and 75/67. There were also pumps installed in different places within the 

Linde Site. In 1996, the pump situated next to Buildings 30 and 31 was surveyed by safety 

personnel using a GM counter. It was found to be contaminated. An outside contractor, William 

Environmental Company (WEC), was brought in to decontaminate the pump. WEC workers 

pulled the pump out without any personal protective equipment. They bare-handedly cleaned the 

pump out. 

The sump in the shower room did not have any filtration system to filter out contaminants. Later 

the sewage piping was found to be contaminated. 

The town of Tonawanda supplied the drinking water for the Linde Site. There were back flows 

onsite at Linde to help prevent water from getting offsite. 

Off-normal Occurrences 

No large incidents were reported by site experts; however, there were small fires periodically. 

The underground tunnels were frequently flooded with rain or snow water. Contaminants from 

buildings, operations, dirt piles, and soils would wash down into this tunnel system without any 

stoppage or barrier. Maintenance workers would regularly go through flood water in the tunnel 

to access utility and equipment. For example, a maintenance crew went down the tunnel to 

replace all the brackets for the 4160 cable line underneath Building 31. They were wearing 

helmets for head protection. During the welding process, the weld exploded and hit the welder 
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on his eyebrow through the helmet. Two weeks later, the wound in the eyebrow of the welder 

was still bleeding. Later on, he was diagnosed with skin cancer. Safety personnel detected 

contamination in the Building 14 tunnel access area in later years. 

In Building 14, the maintenance personnel were working the utility system. A Linde safety 

employee used a Geiger counter and a cutie pie to survey the working areas. The reading went 

off the scale because it was too hot. They notified the MED/AEC office. They sent someone to 

check the area out. But they never heard the results back from AEC or safety. The results of the 

survey were not shared with the employees involved in the incident. 

Maintenance removed roof leaders (5-in pipe to drain the roof) from Building 30. A seam 

grinder was used to cut the pipes. The operations continued during this renovation activity. The 

nuclear physicist came in to perform a survey and told the foreman that the material could not be 

shipped to the scrap file. The pipes were placed on a cart and stored in a corner. The pipes were 

eventually cut down and shipped to the nuclear pile. 

There was a situation where a contractor was brought in to dig up a water line. The individuals 

started digging in the area. Safety came around with a GM counter and indicated that the work 

had to stop due to the presence of radioactivity. There was no hand or foot survey performed 

following exit from the hole. 

Medical Exposure 

The company gave periodic physicals to all employees. During the MED era the physicals were 

more extensive and included: 

• Urinalysis 

• Blood Analysis 

• Chest X-ray (front and side view) 

• EKG 

• Hearing Test 

• Vision Test 

• General Physical 

In the 1940s, workers were given physical examination once a year. For workers working in 

special projects, like uranium separation buildings, they might receive physical examination 

every six months. In the 1950s, medical examination for Linde employees was changed to every 

5 years. For those who worked for special projects, they would still receive medical examination 

every six months. 

If an employee was expected to travel internationally, they also provided the appropriate 

inoculations. In later years, medical examination became age-based. Medical records contain 

the results of physicals and any information on work related injuries and treatment. 

In 1981, the company wanted to expand the physical examination to include pulmonary function 

tests, x-rays, and blood testing. The company would not provide the workers with an 

explanation of why this was necessary. This led workers to ask whether safety conditions were 
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appropriate. These additional tests became optional since the company refused to provide an 

explanation for why the extra tests were necessary. 

Records 

Personnel records were maintained on a Cardex file which provided the date, job title, job code, 

and demographic information. The Cardex file is not directly linkable to a building or area of the 

site. Claimants have had some difficulty obtaining personnel, medical, and/or monitoring 

records from Union Carbide. 

Union Carbide shipped records to a repository (salt mine) in Vermont, which was commonly 

referred to as “the cave.” These records were inventoried prior to shipment. Older records were 

typed on mylar paper which was fragile and crumbling in some cases. Project-related 

documentation was typically kept onsite for 3 years, and then shipped offsite. Some records 

were taken to the incinerator and burned, although the content of these records was unknown. 

The records in Vermont include project-related records, personnel records, and medical records. 

The site eventually started to scan drawings and other records into electronic form. 

Maintenance took copies of paperwork to the job site for reference. They would throw these 

records away after the job. 

The only dose record ever provided by Linde management to a worker was during a Building 14 

incident in 1950s. A worker was exposed to high dose when the door exploded open. 

Chemical Exposure 

Site experts indicated that they had received chemical exposure also. Some of the chemicals 

mentioned during the course of the interviews were silicon rubber, asbestos, molecular sieves, 

ozone, mercury, acid fumes, metal powders (e.g., aluminum, copper, steel), acetone, 

trichloroethylene (TCE), and other chemicals used in the laboratories. There were lime pits 

around Building 30. The company historically dumped TCE to the environment. 

The acid deck was used to clean parts with various chemicals. The company provided special 

protective clothing for work with corrosives and silicon. Acid burns still occurred as a result of 

working with acids. 

Miscellaneous 

During the course of the interview, several individuals indicated that there were numerous 

illnesses and cancer deaths associated with the Linde worker population. One individual had 

been diagnosed with asbestosis. Within the 100 Building, eight (8) individuals (i.e., engineers, 

technicians, clerical workers) developed cancer over a short period of time, although they were 

not routinely assigned to the production areas. 
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Sources of Additional Information: 

The FACTS organization offered SC&A the opportunity to review six boxes of documents 

which included FUSRAP documents, Linde contracts, health and safety records, and job title 

descriptions. SC&A also met with Mr. Francis Amendola, Attorney, to review technical 

documents that his law firm collected for their client regarding Linde claims. Additional 

references or resources recommended by site experts included FUSRAP documents, and worker 

meeting notes. 
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ATTACHMENT 4: ISSUE RESOLUTION MATRIX FOR FINDINGS AND 

KEY OBSERVATIONS 

Table A-4: Issue Resolution Matrix for Linde Findings and Key Observations 

Comment 

Number 

TBD 

Number 

Finding 

Number Issue Number and Description 

SC&A 

Page 

No. 

NIOSH Response Board 

Action 

1 ORAUTTKBS-0025 1 

Issue 1: (5.1.1) Unsupported Assumptions and 

Significant Uncertainties in Information Used – 

SC&A has identified numerous assumption or values 

used in missed dose estimations (both internal and 

external) in the Linde Site Profile that are not either 

supported or adequately supported by explanation, 

available data, technical study, or references. Many of 

these parametric assumptions are made arbitrarily 

without adequate technical basis. In some cases, an 

assumption was made or a value was selected from a 

range of estimated values in order to bound a dose 

parameter that is not entirely justified or explained in the 

document. In other cases, the assumption or value 

selected is not deemed by SC&A as bounding. This is a 

serious flaw that significantly affects the credibility and 

validity of the assigned missed dose estimates in this 

Linde Site Profile. 

38 

2 ORAUTTKBS-0025 2 

Issue 2: (5.1.2.2) Use of Air Concentration Data – The 

use of airborne uranium dust concentration data (air 

concentration) as the sole basis for missed occupational 

internal dose estimation is not defensible and claimant 

favorable, because there are significant uncertainties 

regarding using air concentration data to estimate worker 

inhalation intakes at uranium processing facilities. 

Several technical studies, including the 2003 Y-12 study, 

Practical Use of Personal Air Sampling (PAS) Data in 

the Internal Dosimetry Program at the Y-12 National 

Security Complex (Snapp 2003), and the Nuclear 

Regulatory Commission’s NUREG 1400, Air Sampling 

in the Workplace (Hickey 1993), demonstrate that using 

air concentration data would lead to underestimating the 

worker intakes and subsequently the internal exposures. 

The Y-12 study shows as high as 10 times difference 

(underestimation) between intakes derived from bioassay 

data and intakes derived from air concentration data. 

42 

3 ORAUTTKBS-0025 3 

Issue 3: (5.1.2.3) Urinalysis Data – Using air 

concentration data only, but neglecting urinalysis data, to 

estimate worker inhalation intakes in the Linde Site 

Profile is not in full compliance with 42 CFR 82 

requirements. There are eight sets of urinalysis data for 

over 100 uranium workers in the ORAU Database for the 

period between December 16, 1947, and January 30, 

1950. The air concentration data used in the site profile 

are not complete either and are deemed inadequate (see 

Finding 2). However, NIOSH decided to use these air 

concentration data only for dose reconstruction. This 

approach is not in full compliance with the hierarchy 

approach stipulated in 42 CFR 82. 

45 

4 ORAUTTKBS-0025 4 

Issue 4: (5.1.2.2) Time-Weighted Averages – Timeweighted averages of internal and external exposure 

values contain significant uncertainties and frequently 

fail to capture dose to workers in areas of high uranium 

dust concentration. The site profile uses time-weighted 

calculations to determine average dose values for both 

internal and external pathways. In the internal dosimetry 

section, NIOSH determines the time-weighted average 

42 
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Table A-4: Issue Resolution Matrix for Linde Findings and Key Observations 

Comment 

Number 

TBD 

Number 

Finding 

Number Issue Number and Description 

SC&A 

Page 

No. 

NIOSH Response Board 

Action 

air concentration value of 33 MAC by time weighting the 

air concentration data with average worker exposure 

times and summing to determine daily time-weighted 

average air concentrations by job categories. This 

calculational approach would potentially underestimate 

the average air concentrations for high dose or risk tasks 

that a claimant might have participated in at Linde Site. 

5 ORAUTTKBS-0025 O 

Issue 5: (5.1.2.4) Breathing Rate – The Linde Site 

Profile assumed a breathing rate of 1.2 m3

/hour for 

worker intake. This value implies that workers were 

primarily involved in light exercise during the course of 

the day. A single value may not be consistent with the 

working conditions in the facility, especially during the 

early years of operation, and is inconsistent with other 

NIOSH site profiles, such as Mallinckrodt, Bethlehem 

Steel, Y-12, INL, SRS, and Hanford. 

45 

6 ORAUTTKBS-0025 O 

Issue 6: (5.1.2.5) Ingestion Rate – The Linde Site 

Profile determines the worker ingestion intake by 

multiplying the inhalation intake by 0.2 (20%). Since the 

inhalation intake is estimated by using air concentration 

data, SC&A believes that the NIOSH approach would 

lead to the underestimation of ingestion intake and 

eventual missed ingestion doses for Linde workers. 

45 

7 ORAUTTKBS-0025 O 

Issue 7: (5.1.2.6) Radon Exposure and Concentration

– The Site Profile used the “lowest indoor concentrations 

measured at the Ceramics Plant during African ore 

processing” as the upper limit to both indoor and outdoor 

radon concentrations. The assumed indoor radon 

concentration of 10 pCi/L is based on the lower limit of 

detection. SC&A believes these assumed radon 

concentration values based on the GM of measurements 

are not claimant favorable and representative of the 

actual exposure conditions that the Linde workers 

experienced during the period of operation from 1942 to 

1954. 

46 

8 ORAUTTKBS-0025 5 

Issue 8: (5.1.2.7) Raffinate Trace Radionuclides – The 

dose consequences of raffinate trace radionuclides have 

not been adequately addressed in the Linde Site Profile. 

Raffinate contains Ac-227 and Pa-231, which are in the 

U-235 decay chain, as well as Th-230. Possible doses 

from raffinate-related exposures have not been evaluated 

in the site profile. Inhalation of even small quantities of 

some raffinates, such as filter cake (one of the waste 

products at Linde Site), could result in significant doses 

to the workers. The issue of potential airborne 

contamination of raffinates must be more carefully 

assessed. 

46 

9 ORAUTTKBS-0025 6 

Issue 9: (5.1.2.8) Assigned Work Hours – The number 

of work hours used in calculating occupational internal 

and external doses for workers is inconsistent for 

different periods of Linde operations and, therefore, not 

claimant favorable. The site profile represents in Table 4 

(Davidson 2005, p. 24), and in many other places, that 

workers at Linde had longer workweeks than 40 hours 

per week, and, in some cases, the workweeks were as 

long as 9 hours per day, for 6 days a week and 50 weeks 

per year. But, in most instances, NIOSH uses the 

standard 40 hours per week assumption for the missed 

dose estimation. This approach is not only inconsistent 

but also not claimant favorable. 

47 

10 ORAUTTKBS-0025 7 Issue 10: (5.1.2.9) Surrogate Air Concentration Data

– Using the GM of air concentration data of seven AWE 47 
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Table A-4: Issue Resolution Matrix for Linde Findings and Key Observations 

Comment 

Number 

TBD 

Number 

Finding 

Number Issue Number and Description 

SC&A 

Page 

No. 

NIOSH Response Board 

Action 

facilities in New York from a 1949 AEC/NYOO report 

(AEC 1949a) as surrogate data to develop Linde sitespecific worker inhalation intakes for the entire period of 

Linde Operation from 1942 to 1954, is over-reaching and 

may, potentially, underestimate the missed occupational 

internal dose to workers. This approach is inappropriate 

because the surrogate data are very limited and not 

representative of the actual Linde operation condition 

because, at Linde, ventilation was poor or non-existent, 

and adequate radiation protection practices had not yet 

been developed in the earlier years of operation. 

11 ORAUTTKBS-0025 11 

Issue 11: (5.1.2.10) Use of Geometric Mean Values – 

The statistical analysis approach used in the Linde Site 

Profile is not bounding and, most importantly, not 

claimant favorable. In Table 6 of the Occupational 

Internal Dose Section (Davidson 2005, p. 33), the site 

profile lists the geometrical means or the geometrical 

standard deviation values for measured radon 

concentrations during African ore processing. Firstly, 

there are no supporting calculations or data to show how 

these geometrical quantities are calculated. Secondly, the 

use of geometric means and standard deviations of 

airborne radon concentrations as default values could be 

considered claimant neutral and not claimant favorable. 

Unless there is good reason to believe that a given 

worker was exposed to the full distribution of the 

measured concentrations and could not have experienced 

protracted exposures to higher than average radon 

concentrations, it may be more appropriate to use the 

upper 95th percentile as the default exposure level. 

48 

12 ORAUTTKBS-0025 9 

Issue 12: (5.1.2.11) Lack of Comprehensive 

Uncertainty Analysis – There are no uncertainties or 

potential errors estimated for different assumed 

parameters and factors used in the estimation of 

occupational internal dose in the site profile. An 

assessment of uncertainties, as required by OCAS-IG001 and OCAS-IG-002, has not been adequately 

developed for air concentration and radon measurement 

data used in lieu of the absence of adequate bioassay data 

to assign internal dose. 

48 

13 ORAUTTKBS-0025 8 

Issue 13: (5.1.3) Complex Missed External Dose 

Surrogate System – The Linde Site Profile uses a very 

complex scheme to evaluate missed occupational 

external dose to Linde workers from 1942 to the present 

time. In this scheme, NIOSH/ORAU used a combination 

of film badge data, solid sample analysis results, and 

facility field measurements to estimate missed external 

doses to workers in different periods of the Linde 

operations. These data are, however, limited and, most 

importantly, not facility/building specific. Furthermore, 

the Linde Site Profile uses different sets of data to 

estimate worker beta and gamma doses separately. 

49 

14 ORAUTTKBS-0025 8 

Issue 14: (5.1.3.4) Film Badge Data – The use of the 

1948 weekly film badge data for assigning both beta and 

gamma doses during the removal of equipment in 

Building 30 is not appropriate for the entire period from 

1949 to 1954. These beta and gamma dose assignments 

in Table 36 contain median weekly photon doses and 

weekly median electron doses for use of unmonitored 

workers from 1942 to 1954. These dose assignments are 

not likely to capture the full range of external exposures 

during that time period. Table 36 is hardly representative 

of various facilities and job functions that defined Linde 

58 
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Table A-4: Issue Resolution Matrix for Linde Findings and Key Observations 

Comment 

Number 

TBD 

Number 

Finding 

Number Issue Number and Description 

SC&A 

Page 

No. 

NIOSH Response Board 

Action 

operations and processes. Another problem in Table 36 is 

that some of the beta and gamma doses cannot be 

reproduced or traced back to the original sources. For 

example, there is no explanation or discussion on how 

the 1947 and 1949 (beta/gamma/neutron) doses were 

calculated, since they are all based on 1947–1949 weekly 

film badge data presented in Table 29 and Table 31 of 

the site profile. 

15 ORAUTTKBS-0025 8 

Issue 15: (5.1.3.5) Survey Measurement Data – 

Several sets of survey measurement data were used in the 

Linde Site Profile to calculate the missed beta and 

gamma doses for workers from 1942 to 1954. These 

survey measurements do not cover the entire period of 

Linde operation. SC&A believes that NIOSH should 

improve the use of these film badge data, because 

significant gaps exist for time periods when workers 

were not monitored for external or internal exposure. In 

addition, NIOSH did not evaluate or attempt to evaluate 

the adequacy, uncertainty, and accuracy of these data. 

This further weakens the assigned missed worker beta 

and gamma doses for the Linde workers. 

60 

16 ORAUTTKBS-0025 4 

Issue 16: (5.1.3.6) Time-Weighted Averages – Timeweighted averages of external exposure values contain 

significant uncertainties and frequently fail to capture 

doses to workers in areas of high beta or gamma fields. 

In the external dosimetry section of the Site Profile, 

NIOSH determines the time-weighted average beta and 

gamma radiation dose rates during the standby period 

from 1946 to 1947 by time-weighting the dose rates with 

average worker exposure times and summing to yield 

annual time-weighted average by job category (Davidson 

2005, p. 41). This approach would certainly 

underestimate the dose rates for high-dose or high-risk 

tasks in which a claimant might have participated at the 

Linde Site. 

61 

17 ORAUTTKBS-0025 O 

Issue 17: (5.1.3.7) Contaminated Burlap Bags – 

During the interview in Buffalo, Linde site experts and 

past workers indicated that there were many thousands of 

used burlap bags stacked up in the open bay area behind 

Building 30 (see Attachment 3 of this review report). 

These burlap bags were used for transporting uranium 

ore to the Linde site for processing. After the end of the 

operation period, these contaminated burlap bags were 

stored behind Building 30 awaiting disposal. Many Linde 

workers, operation staff and administrative personnel, sat 

on these contaminated bags during break and lunch 

periods over the period of many years. They definitely 

had been exposed at close distance to beta and gamma 

radiation sources left over in those uranium contaminated 

bags. The Linde Site Profile does not estimate the missed 

beta and gamma doses to workers resulting from sitting 

or standing next to those contaminated burlap bags.

61 

18 ORAUTTKBS-0025 7 

Issue 18: (5.1.3.8) Surrogate External Exposure Data

– The lack of complete film badge data for the period 

from 1942 to 1954 at Linde Site represents a period for 

which the potential for unaccounted beta and gamma 

doses is greatest. NIOSH’s use of pre-cleanup survey 

data for the pre-production period from 1942 to 1943, the 

use of eight solid ore samples data for the period from 

1943 to 1946, the use of a 1-day survey data in six 

locations in Building 30 for the period from 1946 to 

1947, the use of two 1-day pre-cleanup survey data after 

62 
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Table A-4: Issue Resolution Matrix for Linde Findings and Key Observations 

Comment 

Number 

TBD 

Number 

Finding 

Number Issue Number and Description 

SC&A 

Page 

No. 

NIOSH Response Board 

Action 

vacuuming and flushing in Building 30 for 1949, and the 

use of post-decontamination survey data for 1950 is 

complex, over-reaching, inadequately supported, and, 

likely, not claimant favorable. In addition, the use of the 

1948 film badge data collected during the removal of 

equipment in Building 30 for assigning both beta and 

gamma doses for the period from 1949 to 1954 is not 

appropriate, because these data do not account for 

external exposures to contaminated burlap bags, 

contaminated soil, and other contaminated sources during 

the cleanup activities. 

19 ORAUTTKBS-0025 6 

Issue 19: (5.1.3.9) Assigned Work Hours – The Linde 

Site Profile states in Table 4 (Davidson 2005, p. 24) and 

several other sections that workers had longer 

workweeks than the standard 40 hours; as high as 9 hours 

per day, 6 days a week and 50 weeks per year. However, 

in calculating external exposure values, NIOSH uses 

different work-hour values. SC&A believes that applying 

these different work-hour values to the missed 

occupational external dose estimation would 

underestimate the eventual missed dose or exposure 

assignments. NIOSH should use a set of consistent and 

claimant-favorable work hours for use in the dose 

reconstruction. 

62 

20 ORAUTTKBS-0025 11 

Issue 20: (5.1.3.10) Geometric Values – The 

geometrical approach used in the Linde Site Profile is not 

bounding and, most importantly, not claimant favorable. 

In Tables 13, 14, 15, 17, 18, 21, 23, 24, 25, 26, 29, 30, 

31, 32, and 33, the site profile lists the geometrical means 

or the geometrical standard deviation values for various 

assigned default assumptions. Firstly, there are no 

supporting calculations or data to show how these 

geometrical quantities are calculated. Secondly, the 

geometrical approach does not provide maximized 

default values to arrive at claimant-favorable worker 

doses. Thirdly, NIOSH does not provide comparison of 

this geometrical approach with NIOSH-prescribed 95th

percentile values. NIOSH should re-evaluate the 

uncertainties associated with this geometrical approach. 

63 

21 ORAUTTKBS-0025 9 

Issue 21: (5.1.3.11) Lack of Comprehensive 

Uncertainty Analysis – An assessment of uncertainties, 

as required by OCAS-IG-001 and OCAS-IG-002, has not 

been adequately developed for air concentration and 

radon measurement data used in lieu of bioassay data to 

assign internal dose; and, for external exposure data 

(including film badge beta and gamma measurements, 

and survey measurements), used to assign external dose. 

63 

22 ORAUTTKBS-0025 10 

Issue 22: (5.1.4.1) Outdoor Doses – The Linde Site 

Profile does not address missed occupational 

environmental doses to workers. NIOSH did evaluate 

several potential outdoor beta and gamma exposures to 

workers, but, in some cases, NIOSH ignores the outdoor 

doses (Section 4.1.3.1.2, p. 46; Section 4.1.3.2.2, p. 54) 

after the doses are calculated. 

64 

Note: O-Observation