SEC Petition Evaluation Report 

Petition SEC-00088 

Report Rev #: 1 Report Submittal Date: October 18, 2010 

Subject Expert(s): Thomas P. Tomes

Site Expert(s): N/A

Petition Administrative Summary

Petition Under Evaluation

Petition # Petition 

Type

Petition B

Qualification Date

DOE/AWE Facility Name

SEC-00088 83.13 August 17, 2007 Texas City Chemicals, Inc. (TCC)

Petitioner Class Definition 

All laborers who worked in all areas at Texas City Chemical, Inc. from January 1, 1952, through December 31, 

1956.

Class Evaluated by NIOSH

All workers who worked in any area at Texas City Chemicals, Inc., from October 5, 1953, through December 

31, 1956.

NIOSH-Proposed Class to be Added to the SEC

All Atomic Weapons Employer employees who worked at Texas City Chemicals, Inc., from October 5, 1953,

through September 30, 1955, for a number of work days aggregating at least 250 work days, occurring either 

solely under this employment or in combination with work days within the parameters established for one or 

more other classes of employees in the Special Exposure Cohort.

Related Petition Summary Information 

SEC Petition Tracking #(s) Petition Type DOE/AWE Facility Name Petition Status

NONE

Related Evaluation Report Information 

Report Title DOE/AWE Facility Name

NONE

OCAS Health Physicist: Thomas P. Tomes 

Peer Review Completed By: [Signature on file] 10/18/2010 

LaVon B. Rutherford Date

SEC Petition Evaluation Reviewed By: [Signature on file] 10/19/2010

J. W. Neton Date

SEC Evaluation Approved By: [Signature on file] 10/19/2010 

Stuart L. Hinnefeld Date
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

2 of 64 

This page intentionally left blank 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

3 of 64 

Evaluation Report Summary: SEC-00088, Texas City Chemicals, Inc.

This is a revised Evaluation Report by the National Institute for Occupational Safety and Health 

(NIOSH) that addresses a class of employees proposed for addition to the Special Exposure Cohort 

(SEC) per the Energy Employees Occupational Illness Compensation Program Act of 2000, as 

amended, 42 U.S.C. § 7384 et seq. (EEOICPA) and 42 C.F.R. pt. 83, Procedures for Designating 

Classes of Employees as Members of the Special Exposure Cohort under the Energy Employees 

Occupational Illness Compensation Program Act of 2000. 

The revision was made to incorporate a change in the Department of Labor’s (DOL) covered period 

for Texas City Chemicals and to reflect the NIOSH change in its recommendation. Texas City 

Chemicals, Inc. was previously listed as an Atomic Weapons Employer (AWE) facility from 1952 

through 1956, the same period that was specified in the SEC petition. The DOL revised the covered 

period to start on October 5, 1953, which was the first day of operations at the newly constructed 

facility. The revised end date of the covered period was changed to September 1955, which was the 

expiration date of the AEC contract for phosphate leached zone ore development work. The changes 

were based on reports providing additional details on the Atomic Energy Commission work at Texas 

City Chemicals; these reports were not available when the previous Evaluation Report was written. 

This revision also incorporates the requirements of OCAS-IG-004, The Use of Data from Other 

Facilities in the Completion of Dose Reconstructions Under the Energy Employees Occupational 

Illness Compensation Program Act, effective August 21, 2008. Some additional documents and

publications concerning Texas City Chemicals, Inc., (TCC) and the phosphate industry were also used 

in this revision. 

This revision also considers comments received on the original Evaluation Report from S. Cohen & 

Associates, under contract to the Advisory Board on Radiation and Worker Health. 

Some technical changes were made based on the new information and the comments received on the 

previous version. The original Evaluation Report provided a large bounding dose estimate from the 

start of plant operations through the end of the covered period under the assumption of exposure to 

concentrated uranium. The specified covered period has been reduced as well as the period in which 

uranium was extracted. This revision reduces the potential exposure to uranium concentrates from 

extraction operations to the period ending March 1954. Maximum doses during the covered period 

after March 1954 are now based on exposure to phosphate compounds unrelated to uranium recovery 

work, which would have been higher than the doses to workers doing development work with 

leached-zone phosphates. The TCC phosphate plant doses, other than from exposure to radon and its 

progeny, are based on doses received at other phosphate facilities. 

An SEC period is now recommended based on the inability to bound radon exposures with sufficient 

accuracy during the AWE covered period. 

Petitioner-Requested Class Definition

The petitioner(s) for SEC-00088, which qualified on August 17, 2007, requested that NIOSH consider 

the following class: All laborers who worked in all areas at Texas City Chemical, Inc. from January 

1, 1952, through December 31, 1956. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

4 of 64 

Class Evaluated by NIOSH

The petitioner-requested class time period encompassed the years 1952 through 1956, which was the 

previously defined covered period. TCC is now defined as an Atomic Weapons Employer (AWE) 

facility from October 5, 1953, through September 1955. It is covered for residual contamination 

through 1977. 

Based on the revised covered period, NIOSH evaluated potential dose to all workers in any area at 

Texas City Chemicals, Inc., from October 5, 1953, through December 31, 1956. The last fifteen 

months of this period are in the residual contamination period. 

NIOSH-Proposed Class to be Added to the SEC

Based on its full research of the class under evaluation, NIOSH has defined a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes all Atomic Weapons Employer employees who worked at Texas City

Chemicals from October 5, 1953, through September 30, 1955, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or in combination 

with work days within the parameters established for one or more other classes of employees in the 

Special Exposure Cohort. The class under evaluation was accepted (see Section 3.0 below) because 

internal dose from exposure to radon cannot be bounded for employees in the phosphoric acid plant. 

Because there is no Texas City Chemicals employment data to associate individuals with specific 

buildings and work areas, it is not possible to define a class of workers limited to those who worked in 

the phosphoric acid building; therefore, NIOSH cannot bound internal dose to all employees on the 

Texas City Chemicals site. 

Feasibility of Dose Reconstruction

Per EEOICPA and 42 C.F.R. § 83.13(c)(1), NIOSH has established that it does not have access to 

sufficient information to: (1) estimate the maximum radiation dose for every type of cancer for which 

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

member of the class; or (2) estimate radiation doses of members of the class more precisely than an 

estimate of maximum dose. Information available from resources is not sufficient to document or 

estimate the maximum internal dose to members of the evaluated class under plausible circumstances 

during the specified period. 

The NIOSH dose reconstruction feasibility findings are based on the following: 

• Principal sources of internal and external radiation dose for members of the proposed class 

included exposures to technologically enhanced concentrations of naturally-occurring uranium and 

decay products and naturally-occurring thorium and decay products present in phosphate rock. 

• NIOSH finds there are insufficient data to estimate dose with sufficient accuracy from workers 

exposed to radon in the phosphoric acid plant at Texas City Chemicals. NIOSH has found no 

radon monitoring for the facility neither during the AEC period nor of the plant when it was in 

operation. Radon data is available from surveys of the site in the 1980s after the phosphate plant 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

5 of 64 

was shut down. NIOSH also determined that it lacks needed information to model radon 

exposures from processing phosphate rock; dose from processing phosphate rock is only 

applicable to EEOICPA dose reconstructions through September 1955. Although the source term 

is known, available information on the building size, layout, and process activities are insufficient 

to model maximum radon exposures with sufficient accuracy. 

• NIOSH finds that it is feasible to bound occupational external dose from uranium extraction 

operations at Texas City Chemicals using source term and process information. 

• NIOSH finds that it is feasbile to bound occupational internal dose from uranium extraction 

operations based on data from uranium ore concentrate processing at other facilities.

• NIOSH finds that is feasible to bound occupational external and internal dose, other than radon, 

for phosphate workers exposed to technologically enhanced naturally occurring radioactive 

material using data from other phosphate facilities. 

Health Endangerment Determination

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is required because 

NIOSH has determined that it does not have sufficient information to estimate dose for the members 

of the evaluated class.

NIOSH did not identify any evidence supplied by the petitioners or from other resources that would 

establish that the proposed class was exposed to radiation during a discrete incident likely to have 

involved exceptionally high-level exposures. However, evidence indicates that some workers in the 

proposed class may have accumulated substantial chronic exposures through episodic intakes of 

radionuclides, combined with external exposures to gamma and beta radiation. Consequently, NIOSH 

has determined that health was endangered for those workers covered by this evaluation who were 

employed for at least 250 aggregated work days either solely under their employment at Texas City 

Chemicals or in combination with work days within the parameters established for other SEC classes 

(excluding aggregate work day requirements). 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

6 of 64 

Table of Contents

Evaluation Report Summary: SEC-00088, Texas City Chemicals, Inc. ................................................. 3

1.0 Purpose and Scope .......................................................................................................................... 8

2.0 Introduction .................................................................................................................................... 8

3.0 SEC-00088 Texas City Chemicals Class Definitions .................................................................... 9

3.1 Petitioner-Requested Class Definition and Basis .................................................................. 9

3.2 Class Evaluated by NIOSH ................................................................................................. 10

3.3 NIOSH-Proposed Class to be Added to the SEC ................................................................ 10

4.0 Data Sources Reviewed by NIOSH .............................................................................................. 10

4.1 Site Profile Technical Basis Documents ............................................................................. 11

4.2 Technical Information Bulletins and Procedures ................................................................ 11

4.3 Facility Employees and Experts .......................................................................................... 12

4.4 Previous Dose Reconstructions ........................................................................................... 12

4.5 NIOSH Site Research Database .......................................................................................... 13

4.6 Documentation and/or Affidavits Provided by Petitioners ................................................. 13

5.0 Radiological Operations Relevant to the Evaluated Class ........................................................... 13

5.1 TCC Plant and Process Descriptions ................................................................................... 14

5.1.1 Uranium Recovery Operations ............................................................................. 15

5.1.2 Development Work .............................................................................................. 16

5.1.3 TCC Operations Summary ................................................................................... 17

5.2 Exposure Periods ................................................................................................................. 18

5.2.1 Phosphate Plant Operation ................................................................................... 18

5.2.2 Uranium Production Period .................................................................................. 18

5.2.3 Development Period ............................................................................................. 19

5.2.4 Residual Contamination Period ............................................................................ 19

5.3 TCC Source Term ................................................................................................................ 19

5.3.1 Source Term for Uranium Recovery .................................................................... 20

5.3.2 Source Term for the Phosphate Plant ................................................................... 22

5.4 Radiations for TCC .............................................................................................................. 26

5.4.1 Alpha .................................................................................................................... 26

5.4.2 Beta ....................................................................................................................... 26

5.4.3 Neutron ................................................................................................................. 27

5.4.4 Photon ................................................................................................................... 27

6.0 Summary of Available Monitoring Data ...................................................................................... 27

7.0 Feasibility of Dose Reconstruction .............................................................................................. 27

7.1 Pedigree of TCC Data ......................................................................................................... 28

7.1.1 Internal Monitoring Data Review ......................................................................... 28

7.1.2 External Monitoring Data Review ........................................................................ 28

7.2 Internal Radiation Doses at TCC ......................................................................................... 28

7.2.1 Internal Dose from Operation of the Phosphate Plant .......................................... 29

7.2.1.1 Monitoring Data from other Phosphate Plants ..................................................... 29

7.2.1.2 Inhalation and Ingestion Intakes from Phosphates ............................................... 33

7.2.2 Internal Dose from Uranium Recovery ................................................................ 34

7.2.3 Internal Dose during the Development Period ..................................................... 37

7.2.4 Internal Dose from Residual Contamination ........................................................ 37

7.2.5 Radon Exposures .................................................................................................. 39
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

7 of 64 

7.2.5.1 Operational Period ................................................................................................ 39

7.2.5.2 Residual Period ..................................................................................................... 40

7.2.6 Internal Dose Reconstruction Feasibility Conclusion .......................................... 43

7.3 External Radiation Doses at TCC ........................................................................................ 45

7.3.1 External Dose from Operation of the Phosphate Plant ......................................... 45

7.3.2 External Dose from Uranium Recovery ............................................................... 47

7.3.3 External Dose during the Development Period .................................................... 49

7.3.4 Shallow Dose ........................................................................................................ 49

7.3.4.1 Shallow Dose from the Phosphate Plant .............................................................. 49

7.3.4.2 Shallow Dose from the Uranium Recovery Plant ................................................ 51

7.3.4.3 Extremity Dose ..................................................................................................... 53

7.3.5 External Dose from Residual Contamination ....................................................... 54

7.3.6 External Dose from X-ray Examinations ............................................................. 54

7.3.7 External Dose Summary ....................................................................................... 54

7.3.8 External Dose Reconstruction Feasibility Conclusion ......................................... 55

7.4 Evaluation of Petition Basis for SEC-00088 ....................................................................... 55

7.5 Summary of Feasibility Findings for Petition SEC-00088 .................................................. 56

8.0 Evaluation of Health Endangerment for Petition SEC-00088 ...................................................... 56

9.0 Class Conclusion for Petition SEC-00088 ................................................................................... 57

10.0 References .................................................................................................................................... 59

Tables 

Table 4-1: Number of TCC Claims Submitted Under the Dose Reconstruction Rule ........................ 12

Table 5-1: AEC Contracts with Texas City Chemicals ....................................................................... 14

Table 5-2: Relevant TCC Chronology and NIOSH-Designated Evaluation Periods .......................... 18

Table 5-3: Uranium Recovery Plant Radionuclide Ratios .................................................................. 22

Table 5-4: Radioactivity Concentrations in Materials Produced from Florida Phosphates ................ 24

Table 5-5: U-238 and Ra-226 Concentrations Reported by FIPR ...................................................... 24

Table 5-6: Bounding Radionuclide Concentrations in Phosphate Plant Materials ............................. 26

Table 7-1: Airborne Radioactivity Concentrations from Phosphates .................................................. 31

Table 7-2: Dust Concentrations at a Phosphate Plant in Idaho ........................................................... 32

Table 7-3: Intakes from Operation of the Phosphate Plant ................................................................. 34

Table 7-4: Intakes from Operation of the Uranium Recovery Plant ................................................... 37

Table 7-5: Intakes from Residual Contamination ............................................................................... 39

Table 7-6: Residual Radon .................................................................................................................. 43

Table 7-7: Summary of Intakes for Texas City Chemicals ................................................................. 44

Table 7-8: Bounding Radon Exposures at Texas City Chemicals ...................................................... 44

Table 7-9: Maximum External Doses in Phosphate Plants ................................................................. 47

Table 7-10: Photons Dose Rates from Yellowcake ............................................................................... 48

Table 7-11: Summary of External Doses for Texas City Chemicals .................................................... 55

Table 7-12: Summary of Feasibility Findings for SEC-00088.............................................................. 56
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

8 of 64 

SEC Petition Evaluation Report for SEC-00088

ATTRIBUTION AND ANNOTATION: This is a single-author document. All conclusions drawn from 

the data presented in this evaluation were made by the DCAS Health Physicist: Tom Tomes. The 

conclusions were peer-reviewed by the individuals listed on the cover page. The rationales for all 

conclusions in this document are explained in the associated text.

1.0 Purpose and Scope

This report evaluates the feasibility of reconstructing doses for all Atomic Weapon Employer 

employees who worked in any areas at Texas City Chemicals, Inc., from October 5, 1953, through 

December 31, 1956. It provides information and analyses germane to considering a petition for 

adding a class of employees to the congressionally-created SEC.

This report does not make any determinations concerning the feasibility of dose reconstruction that 

necessarily apply to any individual energy employee who might require a dose reconstruction from 

NIOSH. This report also does not contain the final determination as to whether the proposed class 

will be added to the SEC (see Section 2.0). 

This evaluation was conducted in accordance with the requirements of EEOICPA, 42 C.F.R. pt. 83, 

and the guidance contained in the NIOSH Division of Compensation Analysis and Support’s (DCAS)

Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, OCAS-PR-004.1

2.0 Introduction 

Both EEOICPA and 42 C.F.R. pt. 83 require NIOSH to evaluate qualified petitions requesting that the 

Department of Health and Human Services (HHS) add a class of employees to the SEC. The 

evaluation is intended to provide a fair, science-based determination of whether it is feasible to 

estimate with sufficient accuracy the radiation doses of the class of employees through NIOSH dose 

reconstructions.2

 

42 C.F.R. § 83.13(c)(1) states: Radiation doses can be estimated with sufficient accuracy if NIOSH 

has established that it has access to sufficient information to estimate the maximum radiation dose, 

for every type of cancer for which radiation doses are reconstructed, that could have been incurred in 

plausible circumstances by any member of the class, or if NIOSH has established that it has access to 

sufficient information to estimate the radiation doses of members of the class more precisely than an 

estimate of the maximum radiation dose.

Under 42 C.F.R. § 83.13(c)(3), if it is not feasible to estimate radiation doses with sufficient accuracy 

for members of the class, then NIOSH must determine that there is a reasonable likelihood that such 

 

1 DCAS was formerly known as the Office of Compensation Analysis and Support (OCAS).

2 NIOSH dose reconstructions under EEOICPA are performed using the methods promulgated under 42 C.F.R. pt. 82 and 

the detailed implementation guidelines available at http://www.cdc.gov/niosh/ocas.
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

9 of 64 

radiation doses may have endangered the health of members of the class. The regulation requires 

NIOSH to assume that any duration of unprotected exposure may have endangered the health of 

members of a class when it has been established that the class may have been exposed to radiation 

during a discrete incident likely to have involved levels of exposure similarly high to those occurring 

during nuclear criticality incidents. If the occurrence of such an exceptionally high-level exposure has 

not been established, then NIOSH is required to specify that health was endangered for those workers 

who were employed for at least 250 aggregated work days within the parameters established for the 

class or in combination with work days within the parameters established for other SEC classes 

(excluding aggregate work day requirements). 

NIOSH is required to document its evaluation in a report, and to do so, relies upon both its own dose 

reconstruction expertise as well as technical support from its contractor, Oak Ridge Associated 

Universities (ORAU). Once completed, NIOSH provides the report to both the petitioner(s) and to the 

Advisory Board on Radiation and Worker Health (Board). The Board will consider the NIOSH 

evaluation report, together with the petition, petitioner(s) comments, and other information the Board 

considers appropriate, in order to make recommendations to the Secretary of HHS on whether or not 

to add one or more classes of employees to the SEC. Once NIOSH has received and considered the 

advice of the Board, the Director of NIOSH will propose a decision on behalf of HHS. The Secretary 

of HHS will make the final decision, taking into account the NIOSH evaluation, the advice of the 

Board, and the proposed decision issued by NIOSH. As part of this decision process, petitioners may 

seek a review of certain types of final decisions issued by the Secretary of HHS.

3

 

3.0 SEC-00088 Texas City Chemicals Class Definitions

3.1 Petitioner-Requested Class Definition and Basis

The petitioner(s) for SEC-00088, which was qualified on August 17, 2007, requested that NIOSH 

consider the following class for addition to the SEC: All laborers who worked in all areas at Texas 

City Chemical, Inc. from January 1, 1952 – December 31, 1956.

The petitioner provided information and affidavit statements in support of the petitioner’s belief that it 

is not feasible to estimate, with sufficient accuracy, the radiation doses received by workers at TCC. 

NIOSH deemed the information and affidavit statements provided by the petitioner to be sufficient to 

qualify SEC-00088 for evaluation. This information may be summarized as follows: 

Radiation monitoring records for members of the proposed class may have been 

lost, falsified, or destroyed. 

Information regarding monitoring from TCC is unavailable. 

The information and statements provided by the petitioner qualified the petition for further 

consideration by NIOSH, the Board, and HHS. The details of the petition basis are addressed in 

Section 7.4. 

 

3 See 42 C.F.R. pt. 83 for a full description of the procedures summarized here. Additional internal procedures are 

available at http://www.cdc.gov/niosh/ocas.
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

10 of 64 

3.2 Class Evaluated by NIOSH

Based on its preliminary research, NIOSH expanded the petitioner-proposed class to include all 

workers at TCC. NIOSH recognized from the research that there was insufficient information about 

the differences in exposure potential among the various categories of workers at Texas City Chemicals 

to estimate doses for a particular job description. Therefore, NIOSH evaluated maximum potential 

doses received by any worker at the site. The starting date of the time period under evaluation was 

changed to October 5, 1953, because TCC is now considered an Atomic Weapons Employer (AWE) 

facility starting October 5, 1953. 

3.3 NIOSH-Proposed Class to be Added to the SEC

Based on its full research of the class under evaluation, NIOSH has defined a single class of

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes all Atomic Weapons Employer employees who worked at Texas City

Chemicals from October 5, 1953 through September 30, 1955, for a number of work days aggregating 

at least 250 work days, occurring either solely under this employment, or in combination with work 

days within the parameters established for one or more other classes of employees in the Special 

Exposure Cohort. The class under evaluation was accepted (see Section 3.0 below) because internal 

dose from exposure to radon cannot be bounded for employees in the phosphoric acid plant. Because 

there is no Texas City Chemicals employment data to associate individuals with specific buildings and 

work areas, it is not possible to define a class of workers who did not work in the phosphoric acid 

building; therefore, NIOSH cannot bound internal dose to all employees on the Texas City Chemicals 

site. 

4.0 Data Sources Reviewed by NIOSH

NIOSH identified and reviewed numerous data sources to locate information relevant to determining 

the feasibility of dose reconstruction for the class of employees proposed for the SEC petition. This 

included determining the availability of information on personal monitoring, area monitoring, 

industrial processes, and radiation source materials. 

For this revision NIOSH incorporated information received from the DOE and performed additional 

searches for information on Texas City Chemicals and on the phosphate industry in the 1950s; a few

additional industry publications on TCC in the 1950s were found during internet searches and were

incorporated into this revision. 

In June 2008 NIOSH also received copies of court records from Amoco Oil Company vs. Borden, Inc

(Court records, 1989). The records from that court case between the previous and current owners of 

the former Texas City Chemicals property have been reviewed for any information that may be 

applicable to estimating radiation doses to workers from AEC sponsored work. 

The following subsections summarize data sources identified and reviewed by NIOSH. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

11 of 64 

4.1 Site Profile Technical Basis Documents

A Site Profile provides specific information concerning the documentation of historical practices at 

the specified site. Dose reconstructors can use the Site Profile to evaluate internal and external 

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

individual monitoring data. A Site Profile consists of an Introduction and five Technical Basis 

Documents (TBDs) that provide process history information, information on personal and area 

monitoring, radiation source descriptions, and references to primary documents relevant to the 

radiological operations at the site. The Site Profile for a small site may consist of a single document. 

NIOSH has not developed a Site Profile for TCC; however, as part of NIOSH’s evaluation detailed 

herein, it examined the following TBDs for insights into TCC operations or related topics/operations 

at other sites: 

• Technical Basis Document for Atomic Energy Operations at Blockson Chemical, Joliet, Illinois; 

OCAS-TKBS-0002, Rev. 02; November 21, 2007; SRDB Ref ID: 36611.

• Basis for Development of an Exposure Matrix for Blockson Chemical Company, Joliet, Illinois; 

Period of Operation: March 1, 1951 through March 31, 1962, ORAUT-TKBS-0002, Rev. 01; 

June 29, 2004; SRDB Ref ID: 19480. 

• Site Profiles for Atomic Weapons Employers that Refined Uranium and Thorium, Battelle-TBD6001, Rev. F0; December 13, 2006; SRDB Ref ID: 30673.

• Site Profiles for Atomic Weapons Employers that Refined Uranium and Thorium, Appendix BH – 

International Minerals and Chemical Corporation, Battelle-TBD-6001, App. BH, Rev. 0; July 16, 

2007; SRDB Ref ID: 35365.

4.2 Technical Information Bulletins and Procedures

A Technical Information Bulletin (TIB) is a general working document that provides guidance for 

preparing dose reconstructions at particular sites or categories of sites. A procedure provides specific 

requirements and guidance regarding EEOICPA project-level activities, including preparation of dose 

reconstructions at particular sites or categories of sites. NIOSH reviewed the following TIBs and 

procedures as part of its evaluation: 

• OCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, 

Rev. 0, September 23, 2004; SRDB Ref ID 32022. 

• OCAS-TIB-009, Estimation of Ingestion Intakes, Rev. 0, April 13, 2004; SRDB Ref ID: 22397. 

• ORAUT-OTIB-0006, Dose Reconstruction from Occupationally Related Diagnostic X-ray 

Procedures, Rev. 3 PC-1; December 21, 2005; SRDB Ref ID: 20220.

• ORAUT-OTIB-0024, Estimation of Neutron Dose Rates from Alpha-Neutron Reactions in 

Uranium and Thorium Compounds, Rev. 00; April 7, 2005; SRDB Ref ID: 19445.
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

12 of 64 

• ORAUT-OTIB-0043, Characterization of Occupational Exposure to Radium and Radon Progeny 

During Recovery of Uranium from Phosphate Materials, Rev. 00; January 6, 2006; SRDB Ref ID: 

22596.

• ORAUT-OTIB-0070, Dose Reconstruction During Residual Radioactivity Periods at Atomic 

Weapons Employer Facilities, Rev. 00, March 10, 2008; SRDB Ref ID 41603. 

4.3 Facility Employees and Experts

Telephone interviews were conducted with one former TCC employee and a TCC petitioner whose 

spouse worked at the TCC site. The purpose of these interviews was to gain additional information 

and insight into TCC operations during the applicable time period. The questions and responses have 

been documented in the Site Research Database (SRDB). Worker Outreach meetings were also held.

• Personal Communication, 2007a, Personal Communication with Former TCC Employee; 

Telephone Interview by ORAU Team; October 2, 2007; SRDB Ref ID: 35466

• Personal Communication, 2007b, Personal Communication with Survivor of Former TCC 

Employee; Telephone Interview by ORAU Team; October 2, 2007; SRDB Ref ID: 35465 

• NIOSH Worker Outreach Meetings; 2:00 PM and 7:00 PM CDT; October 18, 2007; International 

Union of Operating Engineers Union Hall; 2800 Texas Avenue; Texas City, Texas. Meeting 

minutes are available on the OCAS website (http://www.cdc.gov/niosh/ocas).

4.4 Previous Dose Reconstructions

NIOSH reviewed its NIOSH OCAS Claims Tracking System (NOCTS) to locate EEOICPA-related 

dose reconstructions that might provide information relevant to the petition evaluation. Table 4-1 

summarizes the results of this review for NOCTS data available as of September 28, 2010. 

Table 4-1: Number of TCC Claims Submitted Under the Dose Reconstruction Rule

Description Totals

Total number of claims submitted for dose reconstruction 15 

Total number of claims submitted for energy employees who meet the proposed class definition 

criteria (employment during the period October 5, 1953, through September 30, 1955) 10 

Number of dose reconstructions completed for energy employees who meet the proposed class 

definition criteria 2 

Number of claims for which internal dosimetry records were obtained for the identified years in the 

proposed class definition 0 

Number of claims for which external dosimetry records were obtained for the identified years in the 

proposed class definition 0 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

13 of 64 

NIOSH reviewed each claim to determine whether internal and/or external personal monitoring 

records could be obtained for the employee. No internal or external dose monitoring records were 

found. 

4.5 NIOSH Site Research Database

NIOSH data capture efforts have included internet searches and contacts with the existing site (Amoco 

BP), a DOE representative for exception facilities, Texas State Radiation Group, U.S. Environmental 

Protection Agency (USEPA) Region 6, and the Texas Commission for Environmental Quality. Data 

have also been sought from the Federal Records Center in Fort Worth, Texas. NIOSH has reviewed 

the pertinent data collected by these combined efforts.

NIOSH has also reviewed its Site Research Database to locate documents supporting the evaluation of 

the proposed class. A number of documents in this database were identified as pertaining to TCC

several of which were added since the previous version of the report. These documents were 

evaluated for their relevance to this petition. The documents include historical background on the 

process materials, the industrial process, and Formerly Utilized Sites Remedial Action Program

(FUSRAP)-related residual contamination surveys. 

4.6 Documentation and/or Affidavits Provided by Petitioners

In qualifying and evaluating the petition, NIOSH reviewed the following documents submitted by the 

petitioners:

• Form B with attachment, petition and miscellaneous information from survivor of former TCC 

employee; received March 13, 2007; OSA document id: 102669

• Affidavit, site and medical information from survivor of former TCC employee; received April 4, 

2007; OSA document id: 103975

• Proof of Relationship, affidavit of marriage and employment from survivor of former TCC 

employee; April 17, 2007; OSA document id: 102839

5.0 Radiological Operations Relevant to the Evaluated Class 

This section summarizes operations at TCC and the information available to NIOSH to characterize 

particular processes and radioactive source materials. 

TCC processed phosphate rock that contained naturally occurring radioactive materials (NORM). 

Radionuclides of concern are those that are found in the natural uranium decay series and, to a lesser 

degree, the natural thorium decay series. Chemical operations at the plant resulted in potential worker 

exposures to technologically enhanced naturally occurring radioactive materials (TENORM). NIOSH 

has gathered the source term and process information needed to provide an estimate of the likely 

maximum doses received by workers. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

14 of 64 

5.1 TCC Plant and Process Descriptions 

In 1952 TCC started construction of a plant in Texas City, Texas, to produce animal feed and fertilizer 

from phosphate rock. They contracted with the AEC to simultaneously construct, at TCC’s expense, a 

Uranium Recovery Plant to be used to extract uranium as a byproduct from the phosphates. The 

contract specified terms for sale of the uranium to the AEC. Shake-down operations of the new 

fertilizer plant and Uranium Recovery Plant began on October 5, 1953 (AEC, 1952; AEC, 1953; 

Johnson, 1953). TCC also had a development contract with the AEC (DOE, 1986). 

The TCC plant encountered numerous problems during start-up of the fertilizer plant, and it produced 

a small amount of uranium for the AEC in the first few months of operation; full-scale uranium 

production was never realized due to problems with the fertilizer plant. TCC was operating at a loss

and it ceased operations and filed for bankruptcy in U.S. District Court sometime in 1956. The court 

allowed the Smith-Douglas Corporation to acquire and reorganize TCC and reopen the plant later that 

year. Smith-Douglas did not pursue uranium work with the AEC (AEC, 1953; Greenleaf, 1955; 

Powers, 1979; Corporate Profile, 1958). Smith-Douglas, later acquired by Borden Chemical, operated 

the phosphate plant until its closure in 1977. 

Table 5-1 lists the AEC contracts with TCC as identified in the records of the DOE FUSRAP Program 

(Formerly Utilized Sites Remedial Action Program) (DOE, 1986). 

Table 5-1: AEC Contracts with Texas City Chemicals

AEC Contract Number Dates Description

AT(49-1)-616 Dated February 14, 1952 Letter contract for uranium plant 

construction

AT(49-1)-647

Dated May 12, 1953, lasting 5 years 

from start-up or no longer than 

September 30, 1958

Formal contract to construct plant 

and extract uranium from phosphate

AT(49-6)-910 through September 10, 1955 Process development studies

AT(05-1)-481 Unknown Unknown

Copies of the first two contracts, AT(49-1)-616 and AT(49-1)-647, are available (AEC, 1952; AEC, 

1953). Work conducted under those contracts is discussed in more detail in Section 5.1.1. 

A copy of contract AT(49-6)-910 for process development studies is not available. The FUSRAP 

documentation indicated that the scope of the work under that contract was unknown; however, 

NIOSH received three reports that describe the development work that was performed under the 

contract (Kopf, 1954; Cutter et al., 1954; Greenleaf, 1955). Section 5.1.2 discusses those activities. 

The FUSRAP report also listed contract AT(05-1)-481. There are no details of this contract and 

reports that summarize AEC work at the plant did not include any reference to it. This is discussed 

further in Section 5.1.3. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

15 of 64 

5.1.1 Uranium Recovery Operations 

In February 1952 the AEC signed a letter contract with Texas City Chemicals, Inc. The letter 

contract, which later was superseded by a formal contract, was an agreement between the two parties 

for TCC to construct a uranium recovery unit, at its expense, simultaneous with construction of its 

fertilizer plant, and for the AEC to purchase uranium that was to be produced as a byproduct of the 

fertilizer plant. The contract specified that construction of the plant was to be completed no later than 

October 1, 1953. 

The TCC phosphate plant was designed to have the capacity to process 100,000 tons of Florida 

phosphate rock per year. TCC was to sell all recovered uranium to the AEC. Contract terms limited 

the obligations of both the AEC and TCC to not more than 50,000 pounds of U3O8 annually, specified 

as a concentrate containing at least 50% U3O8 (AEC, 1952; AEC, 1953). The term “U3O8” was used 

by the AEC for inventory purposes to include many uranium feed compounds, not necessarily only 

U3O8. The nominal uranium production capacity based on uranium concentration in the phosphate 

rock source term was about 12 tons per year (Johnson, 1953). Based on concentration of uranium in 

the phosphate rock and the plant capacity, 12 tons could be realized at full plant capacity if complete 

(100%) recovery of uranium was achieved.

According to an internal AEC memorandum (Johnson, 1953), “shake down” operations began at the 

TCC plant on October 5, 1953. “Full-scale” operations were scheduled to begin December 4, 1953, at 

which time the company had planned a ceremony with state and local dignitaries. Subsequent 

documentation indicates that TCC never achieved full-scale uranium production. An AEC Monthly 

Report on Activities of Domestic Production Phosphates (AEC, 1955a), states that TCC operations 

were shut down from January 1954 to the date of the report (December 1955) for “modifications in 

base plant” and that TCC had only produced an “estimated” 303 pounds of U3O8 from “intermittent 

shake-down operations” through December 1953. 

Another AEC report (AEC, c1963) lists details of AEC receipts from all phosphate plants from fiscal 

years 1953 through 1962. For TCC, the report lists a total of 400 pounds of U3O8 produced in fiscal 

year 1954 at a cost to the AEC of $25.00 per pound, which was the maximum price allowed per 

contract terms (AEC, 1952). The report identifies the production period as March 1954. Production 

is listed as zero pounds for fiscal years 1955 and 1956. An adjusting entry of -2 (negative two) 

pounds (and negative $52) is listed for fiscal year 1957. The entry appears to be an accounting 

adjustment and is the final entry for TCC in the report. All subsequent years are listed as 

“Unsuccessful Operation” with no data provided. Final figures for TCC production was 398 pounds 

of U3O8 produced. 

As noted above, the references have some differences in total U3O8 produced by TCC and in the 

months in which it was produced. The AEC 1955 report listed an “estimated” 303 pounds produced 

in December 1953 from intermittent shake-down operations. The more complete report (AEC c1963) 

listed 400 pounds produced in fiscal year 1954, which would include the third quarter of calendar year 

1953. That report separately lists March 1954 as the production period. 

Although the references discussed above have some differences in the exact dates in which uranium 

was produced, they all indicate that TCC produced a relatively small amount of uranium during initial 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

16 of 64 

start-up of the plant. For purposes of this evaluation, 400 pounds of uranium is assumed to have been 

produced intermittently between October 1953 and March 1954. 

The TCC plant was one of four Uranium Recovery Plants built in the early to mid-1950s that was 

designed to recover uranium from phosphoric acid using a solvent extraction process developed by 

Dow Chemical Company. The other three plants were Virginia-Carolina Chemical Corporation, 

International Minerals & Chemicals Corporation (IMCC), and U.S. Phosphoric Products. The TCC 

and Virginia-Carolina uranium extraction plants were never successfully placed into production 

(Wilkinson 1976). 

The production problems encountered by TCC were reported in an AEC memorandum after a visit to 

the TCC plant in January 1955 (Greenleaf 1955). The main subject of the report was the development 

contract work by TCC, but the report also provided information on the uranium recovery contract. 

The report stated that TCC’s fertilizer plant did not function as designed and that the Uranium 

Recovery Plant was not in operation and would not be in operation until the fertilizer plant could 

maintain a reasonably normal level of production. The fertilizer plant had several equipment design 

and performance problems that were so significant that operation of the uranium plant was considered 

hopeless at that time. Before TCC could achieve full fertilizer production it had to order replacement 

equipment. In January 1955 Greenleaf estimated it would be at least three to six months before the 

fertilizer plant would be able to produce enough product to start up the Uranium Recovery Plant. 

There are no additional details on the replacement equipment to get the fertilizer facility fully 

operational, although, as discussed above, AEC production reports indicate the uranium plant was 

never fully operational. 

5.1.2 Development Work 

The AEC had contract AT(49-6)-910 with TCC for development work. That contract expired 

September 10, 1955 (ERDA, 1976; DOE, 1986). TCC submitted reports to the AEC summarizing all 

work performed under contract AT(49-6)-910 (Kopf, 1954; Cutter, et al., 1954). The reports state that 

research was being performed with leached-zone material. The objective was to “develop a cheap 

method of recovering P2O5, Al2O3 and U from this material.” 

Barr described the AEC interest in the leached-zone layer; he also published photographs of four 

plants that were involved in AEC phosphate work at the time, one of which was an aerial photograph 

of Texas City Chemicals. The leached-zone is a layer of earth that overlies the Florida phosphate rock 

“matrix” layer. The matrix layer and leached-zone layer have similar uranium concentrations, but the 

matrix layer has higher phosphate content. Industry utilized the matrix layer as a phosphate (P2O5) 

source while the overlying leached-zone was discarded as waste during mining due to its lower 

economic value. Uranium extraction from phosphate was economically feasible only when it could be 

recovered as byproduct material. Thus, the AEC had contracts with the phosphate industry to find an 

economical method for the chemical industry to utilize leached-zone material so the uranium could be 

recovered as byproduct. Barr identified three approaches being researched. One of the three 

approaches was the separation of P2O5, Al2O3 and U, which was similar to the description of TCC 

development work (Barr, et al., 1955b; Kopf, 1954). 

TCC received a drum of calcined leached-zone material from the Tennessee Valley Authority (TVA) 

laboratory for its initial work. The reports indicate receipts of additional small quantities of leached-
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

17 of 64 

zone material as research progressed. As of the May 1954 report (Cutter, et al., 1954), the 

experiments had not yet started on uranium extraction from the leached-zone material, but TCC

laboratory analyses indicated that the uranium stayed with the phosphate when it was separated from 

alumina. Uranium recovery experiments were planned for the future. 

An AEC official visited TCC on January 12-13, 1955, and wrote a report on the status of uranium 

recovery operations and the leached-zone research activities (Greenleaf, 1955). Greenleaf’s report 

indicates that the leached-zone studies had been suspended sometime prior to his visit and no charges 

were being made to the AEC development contract at that time. TCC was devoting its laboratory 

efforts to more pressing needs concerning trouble with the fertilizer plant, which was not fully 

functional. Up to that point the leached-zone studies had been concentrating on the separation of 

alumina from the phosphate, and estimated that an additional six to eight months of research was 

needed by TCC on alumina phosphate separation. TCC suggested the AEC get another company to 

study some phases of the process due to TCC’s limitations in available resources. Greenleaf’s report 

also mentions that TCC was going to receive some additional 25-pound samples of leached-zone 

material from the TVA facility. 

An AEC Semiannual Report issued in July 1955, for January through June 1955, stated that the AEC 

had completed its studies on recovering uranium from leached-zone materials and concluded it was 

not economically feasible at that time (AEC, 1955b). 

The available reports all indicate that the leached-zone development involved only laboratory-scale 

studies of leached-zone material. A drum of material was initially received and all other quantities 

referenced in the reports identify source material as samples or small quantities. 

5.1.3 TCC Operations Summary

Sections 5.1.1 and 5.1.2 discuss work under three contracts with the AEC. A fourth contract, AT(05-

1)-481, was identified by the DOE FUSRAP program, but no information on the scope of that contract 

is available. The only references to that contract are reports prepared for the FUSRAP project

(ERDA, 1976; ORNL, 1980). However, based on production data and other published reports, no 

additional AEC work is known to have occurred. The development contract AT(49-6)-910 had a 

September 10, 1955, expiration date and the AEC had abandoned research on leached-zone materials 

by July 1955. 

TCC closed the phosphate plant and filed for bankruptcy in 1956. Shortly thereafter, the SmithDouglas company petitioned the U. S. District Court to allow them to reorganize the company and 

reopen the phosphate plant (Corporate Profile 1958). Smith-Douglas was later acquired by Borden 

Chemical. Uranium recovery was not pursued by Smith-Douglass, which operated the plant solely for 

fertilizer and feed production (Powers 1979). 

The Uranium Recovery Plant was torn down (on an unknown date) prior to a FUSRAP program site 

visit in November 1977. Fertilizer production ended at the site in September 1977, and the property 

was sold to the American Oil Company (Amoco) on December 15, 1977 (DOE, 1986; ORNL, 1980). 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

18 of 64 

5.2 Exposure Periods

The TCC AWE covered period for EEOICPA is defined as October 5, 1953, through September 1955, 

with a residual contamination period ending in 1977. All occupationally-derived radiation exposures 

during the covered period must be included in dose reconstructions regardless of whether or not the 

dose was related to AEC atomic weapons-related contract work. After September 1955, only dose 

received from residual contamination that was a result of AEC contract work is considered in dose 

reconstructions (OCAS-IG-003). 

Based on the available information, NIOSH has evaluated the TCC activities and timeframes, and 

designated evaluation periods for dose assessments, as shown in Table 5-2. The periods differ in the 

source terms and potential for exposure, as discussed below. 

Table 5-2: Relevant TCC Chronology and NIOSH-Designated Evaluation Periods

NIOSH-Designated

Evaluation Period Dates Dose

Phosphate Plant Operation Oct. 5, 1953, through Sept. 30, 1955

Industrial dose from 

exposure to TENORM 

during phosphate rock 

processing. Dose 

applicable only during the 

AEC covered period.

Uranium Production Period Oct. 5, 1953, through Mar. 31, 1954 Dose from production of 

400 pounds of uranium.

Development Period Dec. 1, 1953, through Sept. 30, 1955 

Dose from research and 

development with 

phosphate materials.

Residual Contamination Period Oct. 1, 1955 through Dec. 31, 1977

Dose from residual 

contamination from 

production of 400 pounds 

of uranium

5.2.1 Phosphate Plant Operation

Records indicate TCC started shake-down operation of the plant on October 5, 1953, and full-scale 

operations were planned for December 1953. After start-up of the plant, workers were potentially 

exposed to TENORM from chemical processing of phosphate rock. Dose from processing phosphate 

rock is considered applicable throughout the covered period for TCC ending September 1955. Dose 

to workers from operation of the fertilizer and animal feed plant will be evaluated using data from

other phosphate fertilizer plants. 

5.2.2 Uranium Production Period

The October 5, 1953, start-up date of the phosphate plant is presumed to be the start date of potential 

worker exposure to concentrated uranium and associated radionuclides from operation of the Uranium 

Recovery Plant. AEC records indicate that 400 pounds of uranium concentrates were produced 

intermittently between October 1953 and March 1954. For this evaluation a 100-pound per month 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

19 of 64 

production rate is assumed for October 5, 1953 through March 1954, to provide a bounding rate of 

production during this period. 

5.2.3 Development Period

For the purpose of this evaluation, a nominal start date of development work is assumed to be 

December 1, 1953, and work on the contract is presumed to have ended by the September 30, 1955, 

expiration date of the contract; the AEC reported it had completed all its research on leached-zone 

phosphates by the end of June 1955. 

Doses to TCC workers engaged in small-scale phosphate research are presumed to be lower than 

doses to workers who worked in production areas of the animal feed and fertilizer plant discussed in 

Section 5.2.1 because the small scale and limited scope of the development work would not have 

exposed affected workers to the larger source terms associated with production activities. Therefore, 

dose received by the chemists and others involved in research are presumed to be bounded by 

assignment of dose from fertilizer plant operations. 

5.2.4 Residual Contamination Period 

The Residual Contamination Period starts October 1, 1955, and continues through 1977. The only 

doses applicable to EEOICPA dose reconstructions at TCC after October 1, 1955, are the doses 

attributable to residual contamination from AEC work. Uranium extraction operations are assumed to 

have ended not later than March 31, 1954, which is the start date assumed in residual dose 

calculations for the Uranium Recovery Plant, although for dose reconstruction purposes the maximum 

doses from all sources is considered applicable through September 1955. 

5.3 TCC Source Term 

TCC operations began with phosphate rock from central Florida, which was reacted with sulfuric acid 

by the wet process to produce phosphoric acid from which uranium was recovered (Barr, et al., 1955). 

The rock contained naturally occurring radioactive material (NORM), primarily uranium and 

associated decay products, and to a much lesser degree thorium and its associated progeny. An AEC 

memo in 1953 (Johnson, 1953) stated that the phosphate rock used by TCC contained about 0.2 

pounds U3O8 per ton, which is equivalent to about 0.01% U3O8. Blockson reported the average 

uranium content of the central Florida phosphate rock it used in the 1950s was between 0.01% and 

0.014% U3O8 (Lopker, 1951; Stoltz, 1958). Mills, et al. (1977) reported that the marketable rock from 

central Florida had 41 pCi/gm of U-238, which is equivalent to about 0.012% uranium. These values 

are all similar and the differences likely represent the variation that is seen in various batches of 

phosphate rock. A total uranium concentration of 0.014% will be used to bound the average 

concentration in phosphate rock at TCC. 

Chemical processing of phosphate rock at TCC exposed workers to TENORM. The processes

involved with both phosphate chemicals production and with uranium extraction would have altered 

the radioactivity ratios of the various radionuclides. Discussion of the fate of the significant 

radionuclides is provided below. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

20 of 64 

5.3.1 Source Term for Uranium Recovery

U-238 and associated decay products, including Ra-226, are essentially in radioactive equilibrium in 

phosphate rock that has not been chemically processed (Roessler, 1979; FIPR, 1995). During 

production of phosphoric acid by the wet process, phosphate rock is pulverized, mixed with excess

sulfuric acid, and separated into phosphogypsum and phosphoric acid streams. Uranium and radium 

are chemically separated in such a manner that the radium is concentrated in the phosphogypsum 

while the uranium is concentrated in the phosphoric acid (OCAS-TKBS-0002; Guimond, 1975; FIPR 

1995). 

The phosphoric acid produced by TCC occurred in a plant that was built for animal feed and fertilizer 

production. Uranium was extracted as a byproduct from the acid in a separate plant (AEC, 1953). 

Radionuclides in the phosphate rock would be a source term for both the phosphate production areas 

and the Uranium Recovery Plant. 

The distributions of specific uranium and thorium decay chain radionuclides within phosphate source 

materials, and within the various products and waste streams produced by the phosphate ore 

processing industry, have been the subject of numerous studies. While the distributions of 

radionuclides are, in some respects, a function of the specific process, the following generalizations 

can be made for the production of phosphoric acid by the wet process: 

• Radiological equilibrium in the uranium chain appears to be maintained in rock that has not been 

chemically processed (Roessler, 1979; FIPR, 1995). 

• Ra-226 and Po-210 are retained in the phosphogypsum; they do not enter the phosphoric acid 

stream to a significant degree (OCAS-TKBS-0002; Guimond, 1975, page 15; FIPR, 1995, pages 

1-16). 

• Uranium and thorium tend to favor the phosphoric acid phase (OCAS-TKBS-0002; Guimond, 

1975; FIPR, 1995). 

• Since Th-230 is present in the matrix with U-238, it is expected to go into solution along with the 

uranium when leached in sulfuric acid. Th-232, if occupying a different matrix in the mined rock, 

may not be as readily dissolved in sulfuric acid (Coppinger, 1959, page 20). 

• Pb-210 is reported by some references as being retained in the phosphogypsum and by other 

authors as reporting to the phosphoric acid (OCAS-TKBS-0002). 

Uranium and other elements would be present at various stages in concentrations correlated with their 

chemical properties. There are uncertainties with chemical recoveries and potential losses of some 

elements in some of the chemical steps. In lieu of this uncertainty, NIOSH employed the following 

assumptions for isotopic ratios in the production of phosphoric acid from phosphate rock:

• Eighty-five percent of uranium resides in the phosphoric acid (Lopker, 1951; OCAS-TKBS-0002; 

Stoltz 1958). The remainder reports to the phosphogypsum. A higher recovery is considered 

unlikely based on the studies done at Blockson Chemical Company (Blockson). 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

21 of 64 

• Four percent of Ra-226 resides in the acid phase while the remainder reports to the 

phosphogypsum (Hull and Burnett 1996). 

• Although several references indicate the percentage of thorium reporting to the acid is likely to be 

somewhat lower than uranium, this evaluation assumes thorium partitions to acid in the same 

proportion as uranium. The assumption of equal recovery of thorium to uranium in the acid results 

in a higher source term for internal and external dose modeling during uranium recovery (OCASTKBS-0002). If there were more thorium losses to the phosphogypsum stream, the worker doses 

would be lower. 

• The radioactivity ratio of U-238 to Th-232 in the TCC phosphate rock is assumed to be 30:1, the 

same as for Blockson (OCAS-TKBS-0002). This ratio is considered to be a bounding ratio to 

allow for natural thorium and progeny based on reported U-238 and Th-232 concentrations in 

phosphate rock and phosphate products (Mills, et al., 1977; ORAUT-OTIB-0043). Th-232 

progeny are assumed to be in equilibrium. Although most of the Ra-228 likely would have been 

separated and removed with the phosphogypsum, it is assumed to be in equilibrium with Th-232 

for dose-modeling purposes. 

• Pb-210 is assumed to reside in the acid solution in equilibrium with U-238. Various references cite 

data indicating that lead follows the phosphogypsum, while other references report high 

percentages following to the phosphoric acid (OCAS-TKBS-0002). For dose modeling purposes 

and to bound concentrations carrying over to the acid and from ingrowth, Bi-210 and Po-210 are 

assumed to be in equilibrium with Pb-210. 

• All isotopes reporting to the acid are carried through to the drum of dried uranium concentrate in 

the same relative concentration as present in the acid (the uranium concentrate being the highest 

potential source for internal and external dose) (OCAS-TKBS-0002). 

The principal radioactive source term in the Uranium Recovery Plant consisted of dried uranium 

concentrate. The uranium concentration of this material was likely similar to the product produced at 

Blockson, 40% to 60% uranium by mass, although no data are available on the actual concentration 

achieved at TCC. The relative ratios of other radioactive constituents assumed to be present in the 

product are given in Table 5-3. Elevated airborne concentrations of this product material are assumed 

to be present in the Uranium Recovery Plant at the given ratios. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

22 of 64 

Table 5-3: Uranium Recovery Plant Radionuclide Ratios

Radionuclide Relative Ratio1 Notes Normalized to U-2381

U-238 85 Progeny in equilibrium through 

Th-230 1 

U-235 3.87 Progeny in equilibrium 0.0455 

Ra-226 4 Progeny in equilibrium 0.047 

Pb-210 85 Equal to U-238 1 

Bi-210 85 Equal to U-238 1 

Po-210 85 Equal to U-238 1 

Th-232 2.8 Progeny in equilibrium 0.033 

Notes: The data and information contained in this table are from Table 1 in OCAS-TKBS-0002. 1 Ratios given are for progeny without consideration of branching ratios, where applicable.

Wilkinson (1976) reported that the TCC plant was constructed to extract uranium via an organic 

extraction method similar to three plants constructed in Florida, in which UF4 (green salt) was 

produced. However, other references indicate that the uranium concentrate being produced at TCC 

was similar to the Blockson materials. The Fernald operating contractor described the concentrates 

received from both Blockson and TCC as a uranous phosphate. A January 1955 AEC memorandum 

stated that concentrates of Blockson and TCC material were being resampled, and the AEC was 

discussing sampling methods to be used for these two materials in reference to hydroscopity and 

reproducibility of sample results from the two sources of uranium. The AEC made a distinction that 

the chemical characteristics of the Blockson and TCC material were different compared to the green 

salt being produced by Florida phosphate plants (Wunder, 1954; Barr, 1955). A specific description 

of the uranium extraction chemistry at TCC is not available; however, those reports from the AEC 

indicate that the material was not UF4 and instead was similar to the Blockson uranous phosphate. 

The uranium compound produced at TCC is not known with certainty. 

It is assumed that workers were routinely exposed to a drum of uranium concentrate throughout the 

period of uranium recovery operations. Surface contamination of the material is assumed to have 

been present inside the plant during the operational period and to have been in the facility from 

residual contamination after the recovery plant was shut down in 1954. Elevated radon concentrations 

may have been present in the facility from the quantities of Ra-226 that may have carried over into the 

phosphoric acid from the phosphate rock. 

5.3.2 Source Term for the Phosphate Plant

During the AWE covered period dose to workers from exposure to various phosphate compounds at 

the plant must be considered whether or not those exposures were related to work for the AEC. The 

source term in the phosphate plant includes radionuclides in the phosphate rock and phosphoric acid 

discussed in Section 5.3.1, as well as radionuclides that are present in commercial phosphate products 

and wastes generated during production. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

23 of 64 

TCC produced dicalcium phosphate for animal feed. They also produced fertilizer. The available 

references do not include analytical data of the phosphates produced by TCC from 1953 through 1955. 

Information from one former worker suggests that all the phosphate rock had been converted to 

phosphoric acid by the wet process prior to production of products (Personal Communication, 2007a). 

An EEOICPA claimant supplied a copy of an article from the March 29, 1953, edition of the Houston 

Chronicle that discussed the TCC plant that was being constructed. The article described the plans for 

TCC to produce sulfuric acid, phosphoric acid, dicalcium phosphate, ammonia sulfate, and the 

recovery of uranium for sale to the AEC. 

Analytical data from other phosphate plants were used to estimate the concentration of U-238 and 

associated radionuclides in products and wastes at TCC. The data indicates that the concentration of 

U-238 is enhanced, relative to phosphate rock, during production of dicalcium phosphate and 

phosphate fertilizers. Published literature indicates the U-238 concentration and relative radionuclide 

ratios in phosphates vary according to the methods of production. For a given phosphate rock 

concentration, an enhancement factor of two will provide a bounding average U-238 concentration in 

products and wastes at TCC. The available data also includes concentrations of other radionuclides in 

various products. The data is discussed in the paragraphs below. 

Some products, e.g., normal superphosphates, are produced from phosphate rock without a filtration 

step. These are products that are characterized radiologically as having the radionuclides in the same 

approximate radioactivity ratios as found in the phosphate rock. Normal superphosphates and animal 

feed are examples. However, different plants have reported production of animal feed both from 

phosphoric acid and from phosphate rock (Journal, 1953; Menzel, 1968; FIPR, 1998; Gäfvert, 2001). 

TCC used the wet process to produce phosphoric acid. Phosphate products that are produced from 

wet process phosphoric acid involve a filtration step that changes the relative concentration of some 

elements. Products produced from wet-process phosphoric acid are characterized by a much lower 

relative concentration of radium due to a large percentage of it being removed with the 

phosphogypsum waste. Although a relatively large percentage of the radium is found in the 

phosphogypsum, its concentration is not enhanced because the mass of phosphogypsum waste 

produced is larger than the mass of the phosphate rock reacted (FIPR, 1995). 

Concentrated superphosphates, or triple superphosphates (TSP), are produced by reacting phosphate 

rock with phosphoric acid. This product also has an enhanced U-238 concentration, relative to 

phosphate rock. The Ra-226 concentration in TSP is lower than the U-238, but higher than that the 

Ra-226 concentration in ammoniated phosphates. Ammoniated phosphates are produced from 

reacting phosphoric acid with ammonia. The ammoniated phosphates have an enhanced U-238 

concentration relative to phosphate rock, but much lower Ra-226 concentrations (Menzel, 1968; 

Roessler, 1979; FIPR, 1995; FIPR, 1998, Gäfvert, 2001). 

The products produced from each of the methods differ in the ratios and concentrations of the 

radioactive elements. Table 5-4 lists analytical results from common phosphate products. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

24 of 64 

Table 5-4: Radioactivity Concentrations in Materials Produced from Florida Phosphates

Material Ra-226

(pCi/g)

U-238

(pCi/g)

Th-230

(pCi/g)

Th-232

(pCi/g)

Normal Superphosphates 21.3 20.1 18 0.6

Diammonium Phosphates 5.6 63 65 0.4

Concentrated Superphosphates 21 58 48 1.3

Monoammonium Phosphates 5 55 50 1.7

Phosphoric Acid <1 25.3 28.3 3.1

Gypsum 33 6 13 0.3

Reproduced from Mills (1977) Table 2.

Table 5-4, reproduced from Mills (1977), does not have results for dicalcium phosphate animal feed, 

which was produced by TCC. A report from a modern DCP plant in Sweden that imported phosphate 

rock ore from central Florida indicated that all long-lived progeny in the U-238 series were at 

approximate equilibrium in the phosphate rock prior to production of DCP (Gäfvert, 2001). Data 

from the plant indicate most of the uranium from rock ended up in the DCP and had an enhanced U238 concentration of about 120% of that found in the phosphate rock. The Th-230 and Ra-226 

concentrations were very low in the DCP, and they showed up primarily in the sludge, although at 

concentrations noticeably lower than were found in the phosphate rock. The data indicates that the 

radionuclide concentrations in concentrated superphosphates or ammoniated phosphates are higher 

than in the animal feed. FIPR (1998) also reported the most significant concentration of dust came 

from dried processed fertilizers, such as triple superphosphate. 

Analytical data from various phosphate products were reported in the 1998 FIPR report. Table 3 of 

the FIPR report has data from several sources on U-238 and Ra-226 concentrations, which is 

reproduced below in Table 5-5. 

Table 5-5: U-238 and Ra-226 Concentrations Reported by FIPR

Area U-238 (pCi/g) Ra-226 (pCi/g)

Ore matrix 38c 38c

Clays 44a 45a 26c 45b

Sand Tailings 5.3a 5c 7.5a

Rock Concentrate 32c 37c 42d

Rock Pebble 41c 32-41d 42c 37-42

Sodium fluosilicate 0.28d

Gypsum <1c 0.5-6d 26-33d

NSP 21.3b 25d

TSP 57c 20c 21b

MAP 70c 5b 4c 5d

DAP 70c 5.6b 4c

Phosphoric Acid 30c <1c 0.4-0.7d

a Guimond and Windham, 1975

b Guimond, 1978

c Roessler, 1979
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

25 of 64 

d Owen and Hyder, 1980

Notes: NSP = normal superphosphate; TSP = triple superphosphate; 

MAP = monoammonium phosphate; DAP = diammonium phosphate.

The U-238 concentrations for MAP, DAP, and TSP in Table 5-5 were the average values for central 

Florida phosphates reported by Roessler et al., (1979). Additional data reported by Roessler (1979), 

not included in Table 5-5 above, showed the range of U-238 in the ammoniated phosphates varied up 

to 81.8 pCi/g and the TSP values varied up to 72.7 pCi/g. Roessler also reported the Ra-226 

concentrations in filtrate tank sediment varied from 64.6 to 84.1 pCi/g. Ra-226 in the tank scale was 

reported at 384.8 pCi/g. 

The scale is primarily a source of external exposure for workers and maintenance personnel in the 

phosphoric acid filtration area (FIPR, 1998). 

Roessler reported that both the ammoniated phosphates and triple superphosphate had the U-238 

concentrations approaching twice of the concentration in phosphate rock, as supported by the data in 

Tables 5-4 and 5-5. Bounding average uranium concentrations in central Florida phosphate rock will 

be assumed to have been concentrated up to a factor of two at TCC. 

In this Evaluation Report, internal dose from phosphate plant operations at TCC is estimated based 

on bounding total dust concentrations at the plant (see Section 7.2). As stated in Section 5.3 a 

bounding uranium concentration of 0.014% is assumed for TCC phosphate rock. Assuming all 

uranium is U-238, and using a specific activity of 3.4 x 105

 pCi/g, results in a U-238 concentration of 

47.6 pCi/g. U-234 will be assumed in equilibrium with U-238; U-235 is disregarded because the 

radioactivity calculation of U-238/U-234 allow for the small amount of U-235 in the rock. 

Application of a factor of 2.0 for enhanced concentration of U-238 in dry phosphate products results 

in a bounding U-238 concentration of 95 pCi/g. Based on the above data, Ra-226 and Th-230 will 

assumed to be in equilibrium with U-238. This likely overestimates potential concentration of Ra-226 

(relative to U-238), but allows for short term exposure to Ra-226 in filtrate tank sediments. 

Per the discussion in Section 5.3.1, Th-232 (and long lived progeny) is assumed be 3.3% of the 

U-238 concentration. Per Section 5.3.1, Po-210 and Pb-210 behavior in phosphate plants can vary, 

but an assumption of equilibrium with U-238 bounds potential exposure to these radionuclides in the 

phosphate materials that have the greatest dose significance. 

A report from a phosphate plant in Idaho (EPA, 1978) found that Po-210 concentrations were elevated 

relative to U-238 during roasting of phosphate rock in a calciner. However, workers have stated that 

TCC did not roast the phosphate rock prior to processing (NIOSH, 2007a; NIOSH 2007b). 

Bounding concentrations of the radionuclides of significance to internal dose at TCC are provided in 

Table 5-6. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

26 of 64 

Table 5-6: Bounding Radionuclide Concentrations in Phosphate Plant Materials

Radionuclides Concentration (pCi/g)

U-238, Th-230, U-234,

Ra-226, Pb-210, Po-210 95

Pa-231,* Ac-227* 4.3

Th-232, Ra-228, Th-228 3.2

* U-235 progeny assumed in natural abundance relative to U-238. The U-238 and U-234 

concentrations allow for dose from U-235.

Note: Short lived progeny not listed in this table are also assumed to be in equilibrium. 

The values in Table 5-6 are the bounding average concentrations for continuous exposure. Although 

data indicates higher relative concentrations of some radionuclides may be present at certain locations, 

such as high radium content in filtrate tank scale, continuous exposure to bulk products at the 

concentrations in Table 5-6 should provide a bounding source term for the internal exposures 

discussed in Section 7.2. The external dose assessment in Section 7.3 considers dose from Ra-226 in 

tank and pipe scale at TCC. 

5.4 Radiations for TCC

The primary source of radiological exposure from operations performed at TCC for the proposed SEC 

class was naturally-occurring radioactive material (NORM) contained in phosphate rock, primarily 

uranium and thorium, and their associated progeny, and technologically enhanced NORM, or 

TENORM. Potential exposure pathways and sources to be considered include: 

• Internal and external exposure from uranium and associated radionuclides that were extracted and 

concentrated in the Uranium Recovery Plant. 

• Internal and external dose from phosphate rock and manufactured phosphate products. 

• Internal exposure from radon and radon progeny. 

5.4.1 Alpha

Alpha exposure at TCC occurred through intake via inhalation and ingestion. The primary uranium 

isotopes in the phosphate rock include U-238, U-234, and U-235. The most dosimetrically-significant 

associated radionuclides include Th-230, Ra-226, and Rn-222 (radon) and its progeny. Trace amounts 

of natural thorium (and associated progeny) are also present in phosphate rock (FIPR, 1995). 

Personnel exposures to alpha sources are discussed in Section 7.2. 

5.4.2 Beta

Beta exposure sources are discussed in Section 7.3 and include the following: 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

27 of 64 

• Shallow dose from exposure to uranium concentrates. 

• Skin dose from direct contact with material or contamination (e.g., handling and cleaning uranium 

filter media). 

5.4.3 Neutron

Uranium compounds can be a source of neutrons from both spontaneous fission occurring in the 

isotopes of uranium and from alpha-neutron reactions with low atomic number materials, such as 

oxides and impurities. ORAUT-OTIB-0024 describes the expected neutron dose rates from various 

forms of uranium compounds. The neutron dose rates from exposure to the material at TCC are

insignificant compared to the doses assumed to have occurred from other radiation types. 

5.4.4 Photon 

Photon personnel exposure sources are discussed in Section 7.3 and include the following: 

• Exposure to uranium concentrates. 

• Exposure to radium. 

• Occupationally-required medical X-rays. 

6.0 Summary of Available Monitoring Data

NIOSH did not find any personnel or workplace monitoring records for Texas City Chemicals for the 

period under evaluation. Limited site data is available after the phosphate plant closed in 1977. 

7.0 Feasibility of Dose Reconstruction

The feasibility determination for the proposed class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(1). Under that Act and rule, NIOSH must 

establish whether or not it has access to sufficient information either to estimate the maximum 

radiation dose for every type of cancer for which radiation doses are reconstructed that could have 

been incurred under plausible circumstances by any member of the class, or to estimate the radiation 

doses to members of the class more precisely than a maximum dose estimate. If NIOSH has access to 

sufficient information for either case, NIOSH would then determine that it would be feasible to 

conduct dose reconstructions. 

In determining feasibility, NIOSH begins by evaluating whether current or completed NIOSH dose 

reconstructions demonstrate the feasibility of estimating with sufficient accuracy the potential 

radiation exposures of the class under evaluation. If the conclusion is one of infeasibility, NIOSH 

systematically evaluates the sufficiency of different types of monitoring data, process and source or 

source term data, which together or individually might ensure that NIOSH can estimate either the 

maximum doses that members of the class might have incurred, or more precise quantities that reflect 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

28 of 64 

the variability of exposures experienced by groups or individual members of the class as summarized 

in Section 7.5. This approach is discussed in OCAS’s SEC Petition Evaluation Internal Procedures 

which are available at http://www.cdc.gov/niosh/ocas. 

The next four major subsections of this Evaluation Report examine:

• The sufficiency and reliability of the available data. (Section 7.1) 

• The feasibility of reconstructing internal radiation doses. (Section 7.2) 

• The feasibility of reconstructing external radiation doses. (Section 7.3) 

• The bases for petition SEC-00088 as submitted by the petitioner. (Section 7.4) 

7.1 Pedigree of TCC Data

This subsection answers questions that need to be asked before performing a feasibility evaluation. 

Data Pedigree addresses the background, history, and origin of the data. It requires looking at site 

methodologies that may have changed over time; primary versus secondary data sources and whether 

they match; and whether data are internally consistent. All these issues form the bedrock of the 

researcher’s confidence and later conclusions about the data’s quality, credibility, reliability, 

representativeness, and sufficiency for determining the feasibility of dose reconstruction. The 

feasibility evaluation presupposes that data pedigree issues have been settled.

7.1.1 Internal Monitoring Data Review 

NIOSH has been unable to find any record of internal monitoring of TCC workers or any dust or 

radioactivity airborne concentrations for the proposed class. 

7.1.2 External Monitoring Data Review 

NIOSH has been unable to find any record of external dosimetry monitoring of TCC workers or area 

radiation monitoring records for the proposed class. 

7.2 Internal Radiation Doses at TCC

Internal dose from AEC-contracted work is evaluated below, as well as internal dose workers may 

have received from non-AEC work during the covered period. Radiation dose to workers from 

production of chemicals for commercial use is applicable for this evaluation whether or not it was 

directly related to AEC uranium recovery work, but only during the specified AWE covered period 

(OCAS-IG-003). Dose from residual contamination attributable to AEC work is also evaluated. 

Shake-down operations at TCC began October 5, 1953 (Johnson, 1953). From that point on workers 

had the potential to be exposed to technologically enhanced concentrations of radionuclides from 

phosphate rock. There are no monitoring or analytical data available for TCC. Section 7.2.1 

evaluates internal dose and provides bounding intake estimates from operation of the phosphate plant. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

29 of 64 

Internal dose from uranium recovery work for the AEC is considered in Section 7.2.2. Internal dose 

from the AEC development contract is discussed in Section 7.2.3. 

7.2.1 Internal Dose from Operation of the Phosphate Plant

7.2.1.1 Monitoring Data from other Phosphate Plants 

There are no monitoring data from TCC for the AEC covered period; therefore, NIOSH considered 

data from other phosphate plants to assess potential worker intakes. The primary data used by NIOSH 

in the assessment of internal doses at wet process phosphate plants was taken from a 1998 report by 

the Florida Institute of Phosphate Research (FIPR) on dose at Florida phosphate plants that processed 

phosphate rock from central Florida, and a 1978 U. S. Environmental Protection Agency report of 

exposure to workers at an Idaho phosphate plant (FIPR, 1998; EPA, 1978). 

The airborne measurement results from Idaho and the Florida plants were reviewed and high fugitive 

dust concentrations were reported from both studies in certain areas that handled dry phosphate rock 

or dry fertilizers. To estimate intakes of airborne radioactivity at TCC from phosphate plant 

operations, the 50.4 mg/m3

 maximum measured dust concentration at the Idaho plant will be used as a 

bounding maximum average dust concentration. Radioactivity in the dust is calculated using the 

values in Table 5-6. The following paragraphs discuss the Idaho and Florida data and provide the 

rationale for using the Idaho dust concentration. 

The Idaho plant and the Florida plants for which data are available were large plants, processing more 

phosphate rock and producing more phosphate compounds than TCC did in 1953 through 1955. The 

TCC plant was designed to process 100,000 tons of phosphate rock per year (AEC, 1953), equivalent 

to about 2,000 tons per week. A Florida phosphate plant used in the 1998 FIPR report was reported to 

have processed up to 2,000 tons per day (14,000 tons per week) of animal feed or ammoniated 

phosphates alone through the plant’s dry products area. The Idaho phosphate plant referenced in this 

report was a plant in full production and processed one to two million tons of phosphate rock per year 

(20,000 to 40,000 tons per week). TCC was not operating at capacity during the AEC operational 

period, at least through mid-1955 (Greenleaf 1955). The uranium plant was shut down soon after 

startup because the phosphate plant could not maintain sufficient phosphoric acid production due to 

equipment problems. These facts indicate that TCC processed less phosphate rock at a lower rate 

from October 1953 through September 1955 than the Florida and Idaho plants discussed in this report. 

The Idaho and Florida plants are representative of large production facilities that handled large 

quantities of dry rock and phosphate products. 

The chemical processes used in Idaho and Florida are similar to those used at TCC. According to 

publications of phosphate plant operations from the 1950s and from plants in the 1990s, the basic 

process of producing phosphoric acid by the wet process has not changed. Burnet (1957) described 

the methods utilized in the 1950s to produce phosphoric acid. The basic method involves digesting 

finely ground phosphate rock in sulfuric acid in a series of tanks to produce phosphoric acid and 

calcium sulfate (phosphogypsum). The reaction process is dependent on controlling rock particle size, 

acid concentration, reaction time, holding time, and recycle line ratios. The acid and gypsum products 

typically were separated in one of four types of continuous vacuum filters commonly available in the 

1950s. One of those types was the Dorr-Oliver Process, which was the type of filtration used in the 

TCC plant (Corporate Profile, 1958). 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

30 of 64 

The methods described by Burnet (1957) are very similar to methods used in modern phosphate plants 

(FIPR, 1998; FIPR, 1995). Although the basic chemical reactions and processes used in the plants are 

the same, the engineered controls of 1950s-era plants were likely less effective than more modern 

plants. 

FIPR reported a few changes over time in process and practices used at wet process plants. Vacuum 

coolers are used in modern plants to control heat generated from digestion of phosphate rock in 

sulfuric acid. Burnet (1957) stated the method of cooling used in the 1950s was a recycle line in 

which a large amount of unseparated reactants from a downstream digester was fed back into the 

primary digester to cool and control temperature in the initial reaction. That method was replaced by 

the vacuum method in the Florida plants. That design change involves a wet process and has no 

impact on dust concentrations from dry materials in other areas of the facility. 

FIPR also reported that the Florida plants being studied in the 1990s were utilizing a wet method for 

grinding rock that was first developed in 1973. Wet grinding reportedly reduced airborne dust 

emissions by 1,000 pounds per day (reference to the large Florida plants in FIPR 1998). Oil was also 

used on some surfaces in the shipping areas of the Florida plants to control dust. Airborne 

radioactivity concentrations in the ball mill and in some shipping areas were relatively low in the 

Florida Plants indicating the effective use of those dust controls. However, data from some areas 

indicated high levels of fugitive dust. 

FIPR reported the greatest potential for internal dose in the following process areas, in order of 

decreasing dose potential: shipping, rock, dry products, phosphoric acid, and mine (FIPR 1998, p. 

78). The mine category is not applicable to TCC since it did not mine rock. Phosphoric acid 

production is more of a concern for external dose and does not have as much significance for internal 

dose because the handling of liquids and wet materials are the predominate forms of the materials 

used in those areas. Although some intermittent exposure to dried sludge is likely, an assumption that 

the hypothetically maximally-exposed worker was continually exposed to dry phosphate rock or dry 

product results in larger intakes. 

FIPR evaluated air concentrations at Florida phosphate plants in the 1990s (FIPR, 1998). Gross alpha 

and beta concentrations were reported. The lower limits of detection (LLD) for the air sample 

analyses were 1.0 x 10-12 μCi/ml and 1.2 x 10-12 μCi/ml for alpha and beta, respectively. Area 

airborne radioactivity results were recorded for several areas of phosphate rock work. Many samples 

were reported at the detection limits, however, those results are not considered because this evaluation 

considers the areas of highest potential dose. Table 7-1 provides the gross alpha results that were 

reported to be greater than or equal to 2.0 x 10-12 μCi/ml, all of which were in areas that handled 

phosphate rock or dry products; the report provides descriptions of the work in those areas. FIPR 

reported that both high volume air samplers and personal air samplers were used. They selected 

sampling locations based on activities or areas that had the potential to generate dust. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

31 of 64 

Table 7-1: Airborne Radioactivity Concentrations from Phosphates

Plant Area Area Activity Alpha

μCi/ml

D Dry Product, cage mill 2.80E-12

D MAP 2.00E-12

D GTSP 4.20E-12

H Granular 1.59E-11

M Washer 2.20E-12

M Float Plant 2.20E-12

R Rock Receiving, 3rd floor 2.30E-11

R Rock, 3rd floor 3.00E-12

R Rock ball mill 2.00E-12

Reproduced from data in FIPR 1998, Table C-6.

Area key:

D Dry Products

H Shipping

M Mine

R Rock

Only the two highest results will be evaluated.

The single highest airborne alpha radioactivity sample of 2.3 x 10-11 μCi/ml was from the rock 

receiving area. That area of the plant received phosphate rock and processed it for digestion. The 

rock area of the plant would have U-238 and associated radionuclides in the same approximate 

concentration as received in the phosphate rock. The alpha activity measured on the air samples 

were assumed to consist primarily of the long lived radionuclides U-238, Th-230, U-234, Ra-226, and 

Po-210 in equilibrium. Dividing the total alpha concentration by 5 results in an alpha concentration of 

4.6 x 10-12 μCi/ml for U-238 and each of the other four radionuclides. 

The second highest alpha result in the Florida study was 1.59 x 10-11 μCi/ml in the Shipping granular 

area. Granular triple superphosphate (GTSP) has an enhanced U-238 concentration, as discussed in 

Section 5.3.2. Table 5-4 provides a U-238 concentration of 58 pCi/g in concentrated superphosphates. 

The table also provides concentration of other significant alpha emitting radionuclides, but it does not 

include U-234, which would be similar to U-238. Assuming U-234 is equivalent to U-238 at 58 pCi/g 

in concentrated superphosphate, and disregarding relatively insignificant quantities of other alpha 

emitters not listed, the total long lived alpha activity in TSP would be about 186 pCi/g. The U-238 

accounts for 31% (58/186) of that activity. Applying the 31% value to the 1.59 x 10-11 μCi/ml alpha 

result from the Florida data in Table 7-1 results in a U-238 concentration of about 4.9 x 10-12 μCi/ml, 

slightly higher than the U-238 airborne radioactivity estimate for the Rock area discussed in the 

previous paragraph. 

Table 21 of the FIPR report provided concentrations of Ra-226, U-238, and Th-232 from 17 long term 

composite samples of dust accumulation. The U-238 data is used to estimate a range of total dust 

from the 4.9 x 10-12 μCi/ml U-238 air concentration derived from the Granular result in Table 7-1. 

Disregarding four samples identified as gypsum or rock (lower in U-238), the U-238 specific activity 

in the composite samples ranged from 29.11 to 72.85 pCi/g, with a mean of 56.62 pCi/g and a 

standard deviation of 13.17 pCi/g. The average 56.62 pCi/g U-238 concentration is very similar to the 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

32 of 64 

values for Concentrated Superphosphates and TSP in Tables 5-4 and 5-5, respectively. Combing the 

U-238 specific activity with the highest derived U-238 airborne radioactivity of 4.9 x 10-12 μCi/ml, the 

total dust is estimated to be about 87 mg/m3 (range is from 68-169 mg/m3

). The 87 mg/m3 is only a 

nominal value based on a range of likely U-238 concentrations in the highest total alpha result from 

Florida. 

For the Idaho plant, the U.S. Environmental Protection Agency (EPA) reported results of a thorough 

study of dust-loading and radionuclide concentrations in air during phosphate rock processing in 

1975. The phosphate rock used at the Idaho plant did not come from central Florida and had lower 

radioactivity content. But after calcining, the reported U-238 concentration in rock used at the Idaho 

plant was similar to the concentration reported for the rock used in the Florida plants and that reported 

for TCC; the minor differences in radionuclide concentrations should not have an impact on the total 

airborne dust generated. As seen in the Florida data, the data from the Idaho plant indicates that

internal doses are the highest in areas in which phosphate rock and dry phosphate products were 

handled or processed. 

The air samples used in the Idaho study were collected with a two cubic feet per minute portable 

sampler at various locations. Radionuclide concentrations were reported with uncertainties. The 

detection limits were not provided, although some analytical results are reported as less than values. 

The filters were also measured for total dust loading. 

The total dust measurements reported from the Idaho plant, sorted in descending order, are presented 

in Table 7-2. The two highest dust results were reported to be 50.4 mg/m3

 and 33.9 mg/m3 at the 

calciner and the TSP dryer, respectively. 

Table 7-2: Dust Concentrations at a Phosphate Plant in Idaho

Material / Area Dust Loading (mg/m3) 

Calciner 50.4

TSP dryer 33.9

200 Ammophos plant dryer 18.6

Phos. Acid digester 15.3

100 Ammophos plant dryer 8.68

100 plant storage 8.61

Outside control room 6.5

Grinder mill 6.29

Ore unloading, storage 5.43

200 plant storage 4.12

Continuous filter 2.72

Acidulation TSP disch. 2.63

Calciner control room 1.73

Control room 1.5

TSP storage 1.37

Library 0.92
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

33 of 64 

The calciner dust, at 50.4 mg/m3

, would have been fugitive dust from phosphate rock prior to 

chemical processing. The air was sampled for 1 hour, 41 minutes. The reported U-238 concentration

was 1 pCi/m3

 (equivalent to 1 x 10-12 μCi/ml). 

The TSP (triple superphosphate, or concentrated superphosphate) dust, at 33.9 mg/m3, had a slightly 

enhanced U-238 concentration (relative to the calcined ore) based on the isotopic data of the bulk 

material, which was also provided. The air was sampled for 3 hours, 29 minutes. Although the TSP 

dust concentration was lower than the calciner dust, the reported U-238 air concentration was a little 

higher at 1.4 pCi/m3 (equivalent to 1.4 x 10-12 μCi/ml). 

The Idaho plant data also included radioactivity concentrations of other significant radionuclides. The 

concentration and ratio results are a little different that the values reported in Section 5.3.2 for 

phosphates produced from central Florida ores, however, those differences are relatively minor and 

the bounding values used in this evaluation allow for the reported range of concentrations. 

The highest airborne concentrations reported for both the Florida plants and the Idaho plant were at 

levels to which workers are not likely continuously exposed due to the extreme dust loading. 

However, FIPR noted they observed workers had to enter areas near piles of dry product with very 

high dust concentrations and poor visibility. Photographs of phosphate plants are available showing 

localized white clouds during movement of bulk material. 

The bounding concentration for workers at TCC will be based on a limiting average dust 

concentration, which for this evaluation is the 50.4 mg/m3 concentration. The highest reported dust 

concentration cannot be assumed to be an absolute maximum instantaneous dust concentration, 

however, 50.4 mg/m3 likely provides an upper bound average concentration to which a worker would 

have been exposed. An evaluation performed by Wesley R. Van Pelt Associates for the S. Cohen & 

Associates review of the Bethlehem Steel site profile concluded that about 30 mg/m3

 provides a 

plausible upper bound dust concentration for continuous exposure in the work place (S. Cohen & 

Associates, 2005). The air sample at the Idaho calciner was taken over a 1 hour, 41 minute period, 

indicating that some exposures at this concentration were possible, although actual worker exposure to

that concentration in the Idaho plant are not available. 

7.2.1.2 Inhalation and Ingestion Intakes from Phosphates 

A 2,500-hour work-year is assumed. Exposure to 50.4 mg/m3

, at a breathing rate of 1.2 m3/hr, results 

in an annual inhalation intake of 151.2 g of dust per year. The annual total dust inhalation intake is 

multiplied by the 95 pCi/g value for U-238 in Table 5-6 to determination an annual U-238 intake of 

1.44 x 10+4 pCi. Conversion of the result into a calendar-day intake results in a 39 pCi/day chronic 

inhalation of U-238. The other radionuclides identified in the source term in Table 5-6 were 

calculated similarly and the inhalation intakes are provided in Table 7-3 below.

A chronic ingestion mode of intake is assumed to have occurred as a function of the airborne 

radioactivity, assuming a workday ingestion of 0.2 times the amount of airborne radioactivity in one 

cubic meter of air (OCAS-TIB-009). A 300-day work-year was assumed for that calculation resulting 

in a chronic U-238 ingestion of 0.79 pCi per calendar day. Other radionuclides were calculated 

similarly. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

34 of 64 

The bounding intakes for ingestion and inhalation in the phosphate plant are listed in Table 7-3. 

Table 7-3: Intakes from Operation of the Phosphate Plant

Radionuclides Inhalation

pCi/day

Ingestion

pCi/day

U-238, Th-230, U-234,

Ra-226, Pb-210, Po-210

39 0.79

Pa-231, Ac-227 1.8 0.036

Th-232, Ra-228, Th-228 1.3 0.026

7.2.2 Internal Dose from Uranium Recovery

As discussed in Section 5.1, 400 pounds of uranium concentrates were produced between October 

1953 and March 1954. One report indicated 303 pounds were produced through December 1953. For 

this evaluation a 100-pound per month production is assumed for October 1953 through March 1954. 

This rate and time span provides an assumption of 600 pounds uranium concentrate produced. 

Although this is 50% more than the actual production total, it provides a reasonable assurance that the 

timeframe of production is accounted for and that reconstructed doses from uranium recovery are 

favorable.

During the wet chemical processes used to concentrate uranium, contamination and dust exposures are 

minimal. The greatest potential for exposure to radioactive materials associated with the uranium 

recovery process arises in the final packing areas. Here the essentially-pure uranium compound is 

dried and barreled for shipping, resulting in a potentially dusty operation (NRC, 2002a; Eidson, 1984; 

Personal Communication, 2002). The 1953 AEC contract specified that TCC was to produce uranium 

concentrates containing at least 50% U3O8. Similar requirements were found in the contract with 

Blockson. The Blockson uranium recovery process produced concentrates that were reported to 

analyze from 40 - 60% U3O8 (Stoltz, 1958). 

Blockson produced roughly 3,000 pounds of concentrates per month for the AEC, and the workers 

were monitored for uranium intakes. TCC produced less than 100 pounds per month intermittently 

over a few months. A 100-pound per month production rate at TCC would be comparable to a single 

day of production at Blockson, and the uranium source available for inhalation from uranium recovery 

work at TCC would have been only about 3% of the source at Blockson. The Blockson plant is 

known to have required dust collectors for uranium recovery and required housekeeping on each shift 

to reduce dust. Although the source difference can be quantified, unknown differences in the design 

and operation of the plants prevent a direct comparison of intakes for the two plants. 

TCC activities associated with packaging and sampling ore concentrates may or may not have had 

controls such as ventilation and an enclosed drumming station. Additionally, some exposure to dry 

products would have occurred as a result of product sampling, building maintenance, and loose 

surface contamination. 

Inhalation intakes for workers engaged in uranium recovery work at TCC are estimated based on 

worker exposure to airborne uranium concentrates at other facilities. During the 10-year period from 

1948 through 1958, the Health and Safety Laboratory (HASL) of the U.S. Atomic Energy 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

35 of 64 

Commission conducted 60 complete evaluations of occupational hazards in seven uranium refining 

plants. The evaluations consisted of measurements of more than 20,000 individual dust samples. 

These data are summarized in a paper published in November 1960 by the American Academy of 

Occupational Medicine, entitled The Industrial Hygiene of Uranium Refining (Christofano, 1960). In 

this paper, data are presented for various uranium refining operations, including ore handling, ore 

sampling, uranium concentrate sampling, ore digestion, solvent extraction, denitration, oxide 

reduction, hydrofluorination, drum transfer operations, reduction to metal, recasting, fluorination, and 

scrap recovery. 

The AEC uranium concentrate sampling operation described in the paper involved the routine 

handling of 1000-lb. samples of concentrate, 70% to 90% U3O8. Three plants were studied. Specific 

tasks associated with this operation, in order of decreasing exposure, included dumping the 

concentrate (during which airborne gross alpha concentrations ranged from 700 to 4800 dpm/m3

), 

delidding and lidding drums, pipe sampling, and general area. The daily average exposure of workers 

to airborne alpha radioactivity ranged from 90 to 190 dpm/m3. The workers who dumped the ore 

concentrates were exposed without dust controls. Some controls were in place by the AEC in the 

latter period of the study, but the observed concentrations did not improve because of an increase in 

production. 

TCC only produced 400 pounds of uranium concentrates over the lifetime of the operation (less than 

half of a 55-gallon drum). In contrast, the concentrate-handling operations described by Christofano 

involved routine handling of 1000 pound batches of ore concentrates containing 70% to 90% U3O8. 

Exposure to workers handling the ore concentrates at those plants were used to estimate bounding 

exposure to TCC workers in the Uranium Recovery Plant. For this evaluation, NIOSH assumes that 

workers were exposed to the maximum daily average alpha concentration of 190 dpm/m3

 during 

periods of active uranium extraction work. However, the Uranium Recovery Plant at TCC operated 

intermittently for just a few months, so, as previously indicated a favorable asssumption of 100 

pounds of uranium per month is assumed for October 1953 through March 1954. 

As indicated above, a 100-pound per month production rate is assumed for October 5, 1953 through 

March 1954, to provide a bounding rate of production during this period. For this evaluation, a 25-

pound batch is assumed to have been processed every week. Based on a description of how uranium 

concentrates were handled at Blockson, a batch of material may have been in-process over a two-day 

period to allow for filtering, drying and packaging, with the latter two steps being significant for 

potential inhalation. 

Using a 1.2 m3

/hr inhalation rate and a 10-hour work day, a 2-day exposure to 190 dpm/m3 results 

in an inhalation intake of 4,560 dpm, or 2,054 pCi, in the Uranium Recovery Building. The ore 

concentrates reported by Christofano were materials that had been previously processed to remove the 

radium. For this evaluation the alpha activity in those samples is assumed to be 100% uranium. 

Although some non-uranium radionuclides, principally Th-230, may have been present in the alpha 

results reported by Christofano, the assumption of all uranium maximizes the potential uranium 

intake. Assuming U-238 is one half of total uranium, results in a 1,027 pCi U-238 inhalation intake 

for the 2-day period. If workers extracted uranium 2 days per week, they would have been exposed to 

lower levels of airborne radioactivity in the Uranium Recovery Building for the remainder of the work 

week or they would have been exposed to airborne radioactivity from phosphates in other buildings or 

areas the remainder of the work-week. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

36 of 64 

Estimates were made of the airborne uranium in the Uranium Recovery Building on days in which it 

was idle. The potential contamination levels would have been low initially and increased as more 

uranium concentrates were produced, i.e., each batch that was processed is assumed to result in an 

increased potential for airborne radioactivity from residual contamination on subsequent idle days. 

An estimate was made of the surface contamination resulting from the deposition of 190 dpm/m3

airborne uranium. The calculation was made assuming one two-day batch operation per week for 24 

weeks to account for 600 total pounds of uranium as discussed in Section 5.2.2. The suspended 

material is assumed to have settled out of the air with a settling velocity of 7.5 x 10-4 m/s, which is in 

the range of deposition velocities measured in various studies (NRC, 2002b). The settling time for 

each batch was assumed to be 48 hours, although the period during which elevated airborne was 

present was likely less because only part of the operation involved dry materials. Assuming 190 

dpm/m3 for 48 hours for 24 batches, results in a total of 5.9 x 105 dpm/m2 settling on surfaces by the 

time the uranium plant closed in March 1954. It is assumed that none of this had been removed by 

other processes. A resuspension factor of 1 x 10-6/m was applied to the contamination level to 

determine an airborne total uranium concentration of 5.9 x 10-1 dpm/m3

. Thus it is assumed that the 

airborne uranium in the Uranium Recovery Plant on idle days gradually increased to that 

concentration. 

Assuming a 10-hour day, the daily workday intake on idle days in the Uranium Recovery Plant would 

gradually increase to 7.1 dpm, or 3.2 pCi. Table 7-3 lists the bounding U-238 inhalation intake of 

39 pCi/calendar day from phosphate plant work. Converting that to a 6 day workweek results in a 

45.5 pCi U-238 daily workday intake, which is higher than estimated for the Uranium Recovery Plant 

when it was idle. Therefore, during the AEC covered period, intakes from phosphate plant work are 

assumed on days in which uranium was not extracted. This results in a weekly U-238 inhalation 

intake of 1,209 pCi (1,027 pCi from the uranium plant and 182 pCi from the phosphate plant). This 

intake rate would bound intakes from October 5, 1953, through March 31, 1954. 

Ingestion intakes are estimated as a function of the inhalation intakes (OCAS-TIB-009). If the 1,209 

pCi weekly inhalation intake is assumed to have occurred over a fifty-hour period at a breathing rate 

of 1.2 m3

/hr, then the average air concentration could be estimated to be 20.15 pCi/m3. Applying a 

factor of 0.2 results in a workday U-238 ingestion of 4.03 pCi. Assuming a 6-day exposure per week 

results in a weekly U-238 ingestion intake of 24.18 pCi. 

Inhalation and ingestion intakes of other radionuclides were calculated as a function of the U-238 

intake per the respective ratios for the Uranium Recovery Plant and the Phosphate Plant, as discussed 

in Sections 5.3.1 and 5.3.2, respectively. Intakes were normalized to calendar days and the results are 

presented in Table 7-4. The calendar day intake rates in Table 7-4 are applicable from October 3, 

1953, through March 31, 1954. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

37 of 64 

Table 7-4: Intakes from Operation of the Uranium Recovery Plant

Radionuclides

Uranium Plant

Inhalation

(pCi/week)

Fertilizer Plant

Inhalation

(pCi/week)

Combined

Inhalation

(pCi/cal-day)*

Combined

Ingestion

(pCi/cal-day)*

U-238, Th-230, U-234,

Pb-210, Po-210 1,027 182 173 3.5

Pa-231, Ac-227 48.0 8.29 8.0 0.16

Ra-226 48.3 182 33 0.66

Th-232, Ra-228, Th-228 33.9 6.00 5.7 0.11

Notes: Combined intakes are normalized to units of calendar days and represent intakes based an assumption of a 

partial week worked in uranium production and a partial week worked in fertilizer production. 

7.2.3 Internal Dose during the Development Period

As discussed in Section 5.1.2, the Development work done by TCC for the AEC involved laboratoryscale research into methods of separating chemical species of mined phosphate ores and did not 

involve extraction of uranium other than sample analyses. This would have exposed workers to low 

levels of NORM, with essentially the same source term as the phosphate fertilizer plant. The intakes 

of radioactive material by the workers involved in the development work would be expected to be 

lower than the intakes of workers involving handling large quantities of dry material, whether it was 

from uranium extraction or from operation of the animal feed and fertilizer plant. Therefore, dose 

received by workers from AEC-contracted research on phosphate ores would be less than dose 

received by other activities in other areas of the TCC facility. No attempt is made to provide a taskspecific dose estimate because the activities were limited in scope and duration, and the research was 

intermittent according to reports from the AEC (Greenleaf, 1955). 

7.2.4 Internal Dose from Residual Contamination 

Dose from residual contamination in the Uranium Recovery Plant and from residual contamination in 

other areas at TCC is considered. TCC performed laboratory-scale research (Development work) for 

the AEC after the uranium plant was shut down. The materials used for that research did not come 

from TCC’s phosphate plant operations; the AEC furnished small quantities of phosphate materials 

for the research that would result in insignificant doses (see Section 7.2.3). Although TCC continued 

to operate the phosphate plant after the Uranium Recovery Plant closed, the phosphate materials used 

at TCC (excluding AEC Development work) were not used in connection with AEC work, and dose to 

workers from work with non-AEC-related materials are assigned only through September 1955. 

The source term related to AEC activities starting April 1, 1954, would have been from residual 

contamination in the facility. Sources to consider include the Uranium Recovery Plant and other areas 

of the TCC facility in which contamination can be attributed to materials used to produce the 400 

pounds of uranium concentrate for the AEC. 

Section 7.2.2 estimated a bounding intake in the Uranium Recovery Building. The U-238 intake for 

the days in which the plant was not extracting uranium was 3.2 pCi per work-day. The uranium plant 

did not operate after March 1954. This value will be used for the U-238 intake rate from residual 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

38 of 64 

contamination in the Uranium Recovery Building starting April 1, 1954. If a 6-day work-week is 

assumed, it is equivalent to a 2.74 pCi/calendar-day intake rate. 

Intakes from residual contamination from AEC-related activities outside of the Uranium Recovery 

Plant were considered. Uranium recovery for the AEC ended by March 1954, so the phosphate 

materials in process as of April 1, 1954, are unrelated to AEC work. The development work involved 

insignificant source compared to the uranium plant or phosphate plant. 

Section 7.2.1 showed that the highest potential intake from the fertilizer plant was from exposure to 

dry fertilizer products. The U-238 intake estimates from fertilizer in Table 7-3 were derived under the 

assumption that no uranium was removed for the AEC uranium recovery work. That was done to 

provide an estimate of dose from fertilizer during the covered period through September 1955 when 

the uranium plant was idle or shut down. 

Although the materials in process prior to April 1, 1954, would include both AEC related and nonAEC related sources of residual contamination (because the Uranium Plant operated only 

intermittently for the few months), calculations of dose from residual contamination assumes that all 

the material in process prior to April 1954 was AEC-related, with adjustments noted below. 

For AEC-related residual contamination in the fertilizer plant (from processing phosphates used for 

recovering 400 pounds of uranium), the radiological source term available for inhalation would be 

lower because it would have had a significant amount of the uranium and associated radionuclides 

removed and sold to the AEC prior to the phosphates being used for fertilizer. If a nominal 50% 

uranium recovery for the AEC is assumed (lower than the 85% reported for Blockson) and none was 

lost and discarded to phosphogypsum piles, then 50% of the uranium would be available as a 

contaminant for residual contamination on surfaces. Thus when 50% recovery of uranium is assumed 

the residual source term (related to the 400 pounds of recovered uranium) the U-238 source term in 

the phosphate plant is reduced by 50%. This would reduce the bounding U-238 inhalation intake from 

phosphates listed in Table 7-3 to about 20 pCi per calendar day, which will be used as a bounding 

intake rate from residual contamination from the fertilizer plant on April 1, 1954. 

Comparison of the two derived intake rates as of April 1, 1954, 2.74 pCi from the Uranium Recovery 

Plant and 20 pCi from the fertilizer plant, indicates that the intake rate from residual contamination in 

the fertilizer plant is higher than the intake rate estimated from residual contamination in the Uranium 

Recovery Building; therefore, the intake rate from the fertilizer plant will be assumed for residual 

contamination intake calculations starting April 1, 1954. The simplified method used to overestimate 

the residual contamination in the phosphate plant provides reasonable bounding intakes. 

Intake rates from residual contamination after April 1, 1954, is estimated based on consideration of 

depletion factors to account for a gradual reduction of contamination once the AEC work stopped. 

ORAUT-OTIB-0070, Table 3-1, provides depletion factors to apply to derived intakes from residual 

contamination in a facility. Factors are based on a 1%-per-day depletion rate. Favorable factors to 

apply to intake rates are 1.0 for year 1 (no depletion assumed), and 0.03 for year 2. Depletion factors 

from the 3rd year on are held steady at 0.0007 to account for a steady source of contamination. 

Therefore, the U-238 intake rate from residual contamination is 20 pCi/calendar-day starting April 1, 

1954, 0.60 pCi/calendar-day starting April 1, 1955, and 0.014 pCi/calendar-day starting April 1, 1956, 

and continuing through the end of covered employment. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

39 of 64 

The intake rate from residual radioactivity related to AEC uranium recovery work is less than the 

intake rate from on-going fertilizer operations, which are covered exposures through September 1955. 

Therefore, the fertilizer plant intake rates are applied through the end of the TCC’s covered AWE 

period on September 30, 1955. Starting October 1, 1955, intakes from residual contamination apply 

until the end of covered employment. The residual intakes of other radionuclides are applied at the 

same ratios as that applied to intakes from fertilizer operations as done in Table 7-3. Ingestion intakes 

were also estimated using the same methods as in Table 7-3. 

Table 7-5 summarizes the bounding intakes during the residual contamination period beginning in 

October 1955. 

Table 7-5: Intakes from Residual Contamination

Radionuclides Year Inhalation 

(pCi/d)

Ingestion

(pCi/d)

U-238, Th-230, U-234, Ra-226, 

Pb-210, Po-210 Oct. 1, 1955 – Mar. 31, 1956 0.60 0.012

Pa-231, Ac-227 Oct. 1, 1955 – Mar. 31, 1956 0.027 0.0005

Th-232, Ra-228, Th-228 Oct. 1, 1955 – Mar. 31, 1956 0.020 0.0004

U-238, Th-230, U-234, Ra-226, 

Pb-210, Po-210 Apr. 1, 1956 – end 0.014 0.0003 

Pa-231, Ac-227 Apr. 1, 1956 – end 0.0006 <0.0001

Th-232, Ra-228, Th-228 Apr. 1, 1956 – end 0.0005 <0.0001

Notes: Intakes are in units of calendar days. 

7.2.5 Radon Exposures 

The naturally occurring Ra-226 in phosphate rock exposed workers to radon (Rn-222), and short-lived 

radon progeny. Exposures to radon primarily affect the lungs; however, some internal dose to other 

tissues is possible, although relatively insignificant compared to doses from other sources. 

7.2.5.1 Operational Period 

There are no radon monitoring data available for TCC during the AEC operational period or from any 

operating period at TCC. NIOSH considered two methods to estimate radon concentrations at TCC: 

1) using measured data from other facilities; and 2) using source term and process information to 

model potential radon concentrations. 

The radon concentrations at phosphate plants monitored from the 1970s through the 1990s were 

reviewed and reported by the Florida Institute of Phosphate Research (FIPR 1998). The reported data 

vary significantly, with maximum concentrations reported over 70 pCi/L. FIPR measured radon in 

Florida phosphate plants in the 1990s and reported that all areas, other than phosphate rock tunnels, 

have average radon concentration less the EPA guideline (for residences) of 4 pCi/L; the occupied 

work areas were reported to be not significantly different than background radon levels. The 

references indicate workers were not typically exposed to the observed higher concentrations. 

Although the phosphate rock source term is similar, there is not enough information to relate radon 

concentrations in the TCC Acid Plant in the 1950s to radon levels reported at phosphate plants in 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

40 of 64 

more recent years. There is no assurance that the TCC plant in 1953 operated with similar effective 

ventilation as the plants in the 1970s or later. 

NIOSH also considered modeling radon concentrations using source term and process information. 

Under steady-state conditions of production, the average activity concentration of radon in the plant 

can be estimated according to the following equation: 

Vr

SR C ε = (1)

Where:

C = activity concentration of Rn-222 in the plant, Bq/m3

R = processing rate of ore, kg/s 

S = specific activity of Ra-226 in phosphate rock, Bq/kg 

ε = total emanation coefficient of radon from all sources in the plant 

r = outside air exchange rate inside the plant, s-1

V = volume of the plant, m3

As indicated above, building volume is needed to model radon concentration in the phosphoric acid 

plant. The size (volume) of the phosphoric acid plant at Texas City Chemicals is unknown. There is a 

distant overhead picture of the TCC site in the October 1955 issue of Rock Products; it shows several 

buildings, silos, and an open area that appears to be the phosphogypsum pile. The large plant in the 

picture is likely the phosphoric acid plant (they also manufactured sulfuric acid), but estimating the 

building volume based on that distant aerial photograph could, at best, only be done with unknown 

accuracy. NIOSH does not know the layout of the Acid Plant and how it may have been partitioned 

or if radium-bearing phosphates may have also been stored in the building. 

Although NIOSH believes it can reasonably model radon exposures at an operating phosphate facility 

from given source term and other defined parameters, there are significant unknowns in the model 

input parameters at Texas City Chemicals. These include the uncertainty in building volume and 

layout, ore processing rate, and unknown work conditions due to equipment problems. Therefore,

NIOSH concludes that maximum radon exposures cannot be modeled with sufficient accuracy for 

work in the phosphoric acid plant. NIOSH also has no data to determine whether or not a particular 

worker was exposed in the plant. Therefore, maximum radon exposure for all workers at Texas City 

Chemicals who were employed between October 5, 1953, through September 1955, cannot be 

reconstructed. 

Exposure to radon from phosphoric acid production is not covered exposure under EEOICPA after 

September 1955. NIOSH has determined that maximum radon exposures from residual contamination 

related to AEC work can be estimated with sufficient accuracy.

7.2.5.2 Residual Period

The AWE period at Texas City ends in September 1955, which is the end of a contract to do 

laboratory scale development work on leached-zone phosphate ores. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

41 of 64 

Four areas of potential radon exposure were considered to provide an estimate of the maximum radon 

exposures during the residual contamination period. Radon exposures from residual contamination 

were considered in the: (1) laboratory were AEC research was done, (2) the acid plant where 

phosphate rock was digested, (3) the uranium extraction plant, and (4) the phosphogypsum piles. 

The laboratory work involved separation of different chemical species in phosphate ores (P2O5 and 

Al2O3). The radiological source term for the work was very similar to the other phosphate ores used 

in production at the plant, but involved very small quantities in comparison to other activities. Small 

quantities of Ra-226 likely were present as residual contamination in the Uranium Recovery Building 

when it was closed in March 1954, but the Ra-226 primarily was removed with the phosphogypsum 

during phosphoric acid production prior to uranium recovery. Radon exposures that could be 

attributed to materials that were used in AEC related work at TCC during the residual contamination 

period would be primarily from the phosphogypsum pile (or stack) containing the many tons of waste 

generated in producing phosphoric acid. 

There are radon flux monitoring results available from the phosphogypsum stacks at TCC taken from 

February 1983 through September 1984. The results of those tests were summarized by the United 

States District Court for the Southern District of Texas in a 1988 judgment on a lawsuit between the 

current and former owners of the property (Court records, 1989). The combined average flux from 

those measurements was 10.5 pCi/m2

-s. The TCC stacks had been inactive since 1970. The court 

reported that flux measurements from similar inactive stacks in Florida and another plant (location not 

identified) were 4.5 and 4.4 pCi/m2

-s, respectively. Radon gas concentrations at TCC were also 

measured on top of the phosphogypsum stack and near the Administration Building some 200 to 300 

yards from the stack; radon concentrations above background (0.14 pCi/L), were reported to be 0.42 

pCi/L and 0.32 pCi/L, respectively. Radon concentrations at other locations on the TCC property 

were lower.

Court records for the TCC case indicate the measurements were made when the stacks had been 

inactive for a number of years and cited EPA’s findings that radon flux from inactive phosphogypsum 

piles was nearly 5 times less than from active piles due to formation of crust on the pile surface. The 

highest net radon concentration reported at TCC from the 1980s study was on the pile at 0.42 pCi/L. 

Assuming the active pile would have been 5 times higher indicates that a concentration of 2.1 pCi/L 

would have been present when the pile was active. This value compares reasonably well to reports by 

the Florida Institute for Phosphate Research (FIPR 1998). FIPR reported radon results for some 

outdoor areas from Florida plants that had detectable elevated radon concentrations, including 

phosphogypsum stacks (or piles). The results were highly variable and statistics were reported for 5 

locations with elevated results. The median radon concentration for the areas ranged from 1.07 to 

2.72 pCi/L. The 2.1 pCi/L estimate for TCC during periods in which the pile was active should 

provide a reasonable bounding estimate for exposure to radon gas from phosphogypsum at TCC, 

given that workers do not continuously occupy waste piles. However, only a fraction of the 

phosphogypsum at TCC can be attributed to AEC activities.

The amount of phosphate rock that was processed to extract the 400 pounds of uranium that was sold 

to the AEC can be estimated. The central Florida phosphate ores used by TCC reportedly contained 

about 0.01% uranium (Johnson, 1953), although some references have reported slightly higher 

concentrations. If a nominal 50% overall uranium recovery is assumed for TCC’s uranium recovery 

process (lower than the recovery reported by Blockson’s process), phosphate ore containing 800 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

42 of 64 

pounds of uranium was used to produce 400 pounds of uranium for the AEC through March 1954. At 

a concentration of 0.01% and a recovery of 50%, 8 x 106

 pounds, or 4,000 tons, of phosphate rock 

would be required to recover 400 pounds of uranium. If this quantity was spread out over 4 months of 

intermittent operation (full plant production started in December 1954), 1,000 tons of phosphate rock 

was processed per month, presuming uranium had been recovered from all the phosphate rock that 

was processed during that period. 

At capacity TCC could have processed over 8,000 tons phosphate rock per month, although the plant 

was not operating at capacity due to equipment problems. They were operating at less than capacity 

as late as February 1955 (date of last known AEC documented site visit) because new equipment had 

not yet arrived. 

If it is assumed that TCC processed phosphate rock at no more than 1,000 tons per month from 

December 1953 through the end of the AEC period in September 1955, about 22,000 tons of 

phosphate rock would have been processed, of which approximately 4,000 tons, or 18%, was related 

to AEC work. 

According to a brief filed with the United States Court of Appeals for the Fifth Circuit in 1989, the 

former Texas City Chemicals site had over one million tons of radioactive waste stored on site, most 

of it in the phosphogypsum piles that encompassed 35 acres of the 114 acre site. The phosphogypsum 

waste was from production of phosphoric acid from the days of ownership by Texas City Chemicals, 

Inc., through about 1970. According to the court papers phosphoric acid production ended in about 

1970; the phosphogypsum stacks were inactive from 1970 until the plant was closed in 1977 and 

remained inactive after plant closure (Court Records, 1989). 

Based on the reported capacity of the plant, one million pounds of waste is a plausible phosphogypsum mass over the years in which the phosphogypsum stacks were active (FIPR, 1995). Using 

one million tons as an estimate of total phosphogypsum waste in 1970, the estimated 4,000 tons 

related to AEC work represented 0.4% of the total phosphogypsum when the stacks were closed. 

For calculation purposes, phosphogypsum is assumed to have been added to the piles at a steady rate 

starting in 1957. TCC shut down and filed bankruptcy in early 1956 and was reopened in late 1956 

under new ownership. The September 1955 estimate for 22,000 tons of total waste is assumed for the 

time when TCC closed and filed bankruptcy in early 1956. The amount of waste is assumed to have 

increased at a constant rate of 69,857 tons per year from 1957 and reached one million tons by the end 

of 1970. 

Based on the above discussion, a 2.1 pCi/L radon concentration is assumed for exposure to the active 

phosphogypsum pile throughout the period under evaluation. The estimated concentration attributable 

to the AEC at the start of the year is assumed for the entire year. Thus, 0.38 pCi/L (2.1 pCi/L * 0.18) 

is assumed for all of 1957. Table 7-6 below shows the estimated bounding radon concentrations from 

AEC-related waste in each year. The residual contamination period for TCC is listed as ending in 

1977, which is before the date of the radon measurements, therefore, radon is assumed to remain 

constant at the estimated 1971 levels until the end of covered employment. 

Although the waste piles are not typically occupied, it is assumed a worker may worked routinely 

in the vicinity of the piles, which should provide an estimate of the maximum radon exposure 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

43 of 64 

attributable to AEC work. A 0.4 equilibrium factor is assumed to convert pCi/L radon gas to working 

level months (WLM) based on a 2500 hour work year and 1.0 working level (WL) per 100 pCi/L. 

The 1955 exposure is prorated for exposure starting October 1, 1955. The radon gas concentration 

and annual exposures for AEC residual radon is provided in Table 7-6. 

Table 7-6: Residual Radon

Year Total waste, tons AEC waste, tons Fraction AEC AEC radon, pCi/L AEC WLM

1955 22,000 4000 0.18 0.38 *0.0056

1956 22,000 4000 0.18 0.38 0.022

1957 22,000 4000 0.18 0.38 0.022

1958 91,857 4000 0.044 0.091 0.0054

1959 161,714 4000 0.025 0.052 0.0031

1960 231,571 4000 0.017 0.036 0.0021

1961 301,429 4000 0.013 0.028 0.0016

1962 371,286 4000 0.011 0.023 0.0013

1963 441,143 4000 0.009 0.019 0.0011

1964 511,000 4000 0.008 0.016 0.0010

1965 580,857 4000 0.0069 0.014 0.0009

1966 650,714 4000 0.0061 0.013 0.0008

1967 720,571 4000 0.0056 0.012 0.00069

1968 790,429 4000 0.0051 0.011 0.00063

1969 860,286 4000 0.0046 0.010 0.00057

1970 930,143 4000 0.0043 0.0090 0.00053

1971 1,000,000 4000 0.0040 0.0084 0.00049

* WLM values for 1955 have been calculated to include only exposures received from October 1, 1955, through 

December 31, 1955. WLM calculations assume a 2,500 hour exposure for a full work year at the given concentration 

and a 0.4 equilibrium factor.

7.2.6 Internal Dose Reconstruction Feasibility Conclusion 

This evaluation concludes that maximum radon exposures from phosphate plant operations at TCC 

cannot be reconstructed with sufficient accuracy. Dose from phosphate plant operations are not 

covered past September 30, 1955. Other than radon exposures, NISOH concludes that internal doses 

can be reconstructed during the AEC covered period. 

During the residual contamination period beginning October 1, 1955, NIOSH concludes that 

maximum internal and external doses from all sources can be reconstructed. 

Internal dose reconstruction, other than radon, for members of the evaluated class is feasible based on

the use of source-term information at TCC and the use of measured doses of workers from other wet 

process phosphate plants. The above evaluations indicate that internal exposures, other than radon, 

can be bounded for both the operation of the uranium recovery plant and operation of the animal feed 

and fertilizer plant. Based on the claimant-favorable evaluations, the potential internal doses from the 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

44 of 64 

uranium recovery plant bounds dose for the period in which the plant operated. Dose from fertilizer 

operations were higher when the uranium plant was idle and when the AEC work was limited to 

leached-zone phosphate development work. A summary of the bounding intakes and internal 

exposures for all periods at TCC is provided in Tables 7-7 and 7-8. 

Table 7-7: Summary of Intakes for Texas City Chemicals

Time Period Radionuclides Inhalation 

(pCi/d)

Ingestion

(pCi/d)

October 5, 1953, through 

March 31, 1954

U-238, Th-230, U-234, Pb210, Po-210 173 3.5

Pa-231, Ac-227 8.0 0.16

Ra-226 33 0.66

Th-232, Ra-228, Th-228 5.7 0.11

April 1, 1954, through 

September 30, 1955 

U-238, Th-230, U-234, 

Ra-226, Pb-210, Po-210 39 0.79

Pa-231, Ac-227 1.8 0.036

Th-232, Ra-228, Th-228 1.3 0.026

October 1, 1955, 

through

March 31, 1956 

U-238, Th-230, U-234, 

Ra-226, Pb-210, Po-210 0.60 0.012

Pa-231, Ac-227 0.027 0.0005

Th-232, Ra-228, Th-228 0.020 0.0004

April 1, 1956, 

through

the end of covered 

employment

U-238, Th-230, U-234, 

Ra-226, Pb-210, Po-210 0.014 0.0003

Pa-231, Ac-227 0.0006 <0.0001

Th-232, Ra-228, Th-228 0.0005 <0.0001

Notes: Values are normalized for a 365-day calendar year. All intakes are assumed to be 

bounding and applied as constants.

Material solubility should be based on claimant-favorable types available in ICRP Publication 68 (ICRP 

1994).

Table 7-8: Bounding Radon Exposures at Texas City Chemicals

Time Period Component Annual Distribution

October 5 1953, through 

September 30, 1955 Cannot be reconstructed.

October 1, 1955, through

the end of covered 

employment

Radon progeny 

exposure

See Table 7-6 Constant value

Radon gas exposure See Table 7-6 Constant value
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

45 of 64 

7.3 External Radiation Doses at TCC

The principal source of external radiation doses for members of the proposed class was exposure to 

uranium, thorium, and their associated progeny. Because phosphate rock contains NORM, any work 

with this material potentially exposes workers to radioactivity. At TCC, uranium was extracted from 

phosphate rock and concentrated for use by the AEC. Employer-required medical X-ray examinations 

are another potential source of external radiation dose at TCC. External doses received by workers in 

the evaluated class from production of commercial phosphate products are also considered during the 

AEC covered period. 

There are no dosimetry data or radiation measurements available for TCC. Therefore, NIOSH has 

considered worker doses using source-term information and measured doses of phosphate plant 

workers according to OCAS-IG-004, The Use of Data from Other Facilities in the Completion of 

Dose Reconstruction Under the Energy Employees Occupational Illness Compensation Program Act. 

The selection and use of modeled and measured data from other facilities are discussed below. 

There have been numerous reports and studies on doses received by workers at phosphate plants. The 

most extensively published works are reports from the Florida Institute of Phosphate Research (FIPR), 

although the U. S. Environmental Protection Agency (EPA) and others have also published data on 

worker doses. A 1998 FIPR report contains a summary of some of the data that had been previously 

reported in publications by the EPA, NCRP and others (FIPR 1998). Published data indicate average 

annual external doses received by workers at phosphoric acid and fertilizer plants are relatively low. 

FIPR found that it is extremely unlikely for a worker to exceed the 5,000 mrem per year total effective 

dose equivalent annual limit (TEDE) (FIPR 1998, p. 3). External dose is one component of the 

TEDE. The report was not specific to plants operating in the 1950s or in Texas, but the report 

presented data from studies as far back as 1977. The data considered in this Evaluation Report were 

from various phosphate plants that used similar phosphate rock to produce phosphoric acid by the wet 

process. Application of the data to TCC is discussed in the sections below. 

7.3.1 External Dose from Operation of the Phosphate Plant

Shake-down operations of the fertilizer and animal feed plant began October 3, 1953, which is the 

start date assumed for workers being exposed to technically enhanced naturally occurring radioactive 

material (TENORM). Worker doses from phosphate plant operations are considered throughout the 

covered period ending September 1955. Potential dose to workers in this evaluation is based on 

estimates of dose received by workers in wet process phosphate plants. 

TCC was producing animal feed and fertilizer. Phosphoric acid was being produced by the wet 

process (Barr, et al., 1955). The particular wet process used by TCC was known as the Dorr-Oliver 

Process (Corporate Profile 1958). The process involved adding an excess of sulfuric acid to ground 

phosphate rock to produce phosphogypsum and phosphoric acid. It utilized a series of digester tanks 

(attack tanks) followed by removal of phosphogypsum from the acid solution via a traveling pan filter 

or a horizontal rotary filter (Burnet, 1957). Former workers have provided descriptions of the acid 

plant digesters and described the filter as a large vacuum filter table (NIOSH, 2007). 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

46 of 64 

Various publications have reported external photon doses from workers producing various products at 

wet process phosphoric acid plants. Ra-226 is a significant source of external dose in phosphate 

plants and has been reported to concentrate in filtrate tank sediments and scale inside piping (FIPR 

1998, p.10). Higher than normal radiation levels from elevated Ra-226 concentrations have been 

found inside filter pans and inside piping used for transporting weak acid from filter washing. 

A University of Florida study, as reported by FIPR, found gamma radiation levels in residues ranging 

from “100 to 1000 mR/hr;” however, it reported time-weighted average exposures were less than 25% 

of the standards for occupational exposure, although no actual dose values were provided. A 1980 

Department of Energy Study reported average gamma exposure to phosphate workers at less than 0.03 

mR/hr ± 70% (FIPR 1998, p. 11). That rate would result in an exposure of 60 mR per year over a 

2,000-hour work year. The reported uncertainty indicates that the 95th percentile exposure would be 

about 130 mR per year. FIPR also reported results from a 1980s study by Keaton. That was a study 

of dose to subcontractor maintenance personnel who routinely worked on filtration systems at 

phosphate plants. 

The highest dose rates were found on the inside surfaces of pipes used to transport acid wash. The 

acid wash is a weak acid produced by washing the gypsum with water after the phosphogypsum had 

first been separated and washed with acid (return acid). Barium sulfate formed in the gypsum matrix 

plates out in the piping; the buildup can be significant in the weak acid wash lines because it is very 

insoluble in water. The dose rates are attributable to Ra-226 that replaces barium ions in the barium 

sulfate crystals. The amount of scale is a function of product throughput and was reported to be a 

condition that slowly builds over time. When pipes are cleaned with acid to remove the scale, data 

indicate the scale gradually builds back. Worker doses from those sources were reported to be 

effectively controlled by cleaning the scale from equipment every one to three years. Studies were 

performed of plant workers who worked with and around the filtration systems. Jobs included 

cleaning and replacing piping, changing filter cloth and sampling. Dose rates, job durations and 

frequency of work were considered. Dose to several categories of workers were reported; the highest 

exposed worker dose was estimated to be less than 0.220 rem per year (Lardinoye, 1982). 

Workers at TCC from October 1953 through September 1955 were unlikely to have received doses as 

high as those received in fully operational plants that had been operating for a few years because the 

TCC plant was new, and upon startup it would not have the buildup of scale that is associated with 

production. The buildup of scale at TCC would also have been slow due to reduced production 

because of the failure of equipment to perform as designed. Although the plant was operating, it was

not producing enough phosphoric acid to keep the Uranium Recovery Plant open. 

The 1998 FIPR report summarized some of the exposures provided by others and presented results of 

new studies that were performed to better characterize the exposure of various groups of workers at 

phosphoric plants in Florida. Results were obtained from the deployment of LiF and aluminum oxide 

carbon dosimeters. The dosimeter deployment included concurrent monitoring with both dosimeter 

types to achieve optimal sensitivity and the most accurate estimate of deep dose equivalent. Of the 

seven categories of monitored workers, FIPR reported that the geometric mean of dose to personnel 

who worked in the phosphoric acid production areas and in rock handling areas received the highest 

external dose (FIPR 1998, p. 73). Other monitored areas were: dry product, shipping, mine, service, 

and general area. Appendix C of the FIPR report contains the dosimeter data for each category of 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

47 of 64 

workers. The highest dose in each particular category is listed in Table 7-9 below. Reported results 

are annualized deep dose equivalent in mrem. 

Table 7-9: Maximum External Doses in Phosphate Plants 

Job Type or Area Highest annual dose, mrem

Dry Product 210

Shipping 180

Mine 186

Phosphoric Acid 172

Rock 141

Service 184

Highest result for each category from FIPR (1998), Tables C1 through C4. 

Consideration was also given to an evaluation of worker exposures reported by ORAU in ORAUTOTIB-043, Characterization of Occupational Exposure to Radium and Radon Progeny During 

Recovery of Uranium from Phosphate Materials. The upper-bound dose for phosphate plant workers, 

outside of uranium recovery operations, was estimated to be 220 mrem/year, which is the same as 

reported by Lardinoye (1982) and only slightly higher than the maximum dose observed in the FIPR 

study. 

For purposes of this Evaluation Report, a 220 mrem/year bounding external whole-body photon dose 

rate is assumed for work in the TCC phosphate plant after shake-down operations started October 5, 

1953, which is equivalent to 0.00060 rem/calendar day. Dose from operation of the Uranium 

Recovery Plant is considered separately. 

7.3.2 External Dose from Uranium Recovery

The Uranium Recovery Plant produced uranium intermittently upon startup of the fertilizer plant in 

October 1953. The problems with the fertilizer plant led to the shutdown of the uranium plant. As 

discussed in Section 7.2.2, the intermittent uranium extraction from phosphoric acid is assumed to 

have ended by the end of March 1954. External dose to workers during the operation of the uranium 

plant is based on modeled external doses from a drum of uranium product. Exposure to contaminated 

surfaces and air immersion were considered, but those doses are insignificant in comparison to the 

modeled dose received from exposure to bulk material. An assumption of working around bulk 

material provides a larger source term for exposure. 

A bounding dose rate for exposure to uranium concentrates at TCC has been modeled similar to what 

was done for Blockson (OCAS-TKBS-0002); however, differences in the production levels and period 

of operation have been taken into consideration for TCC. Blockson produced roughly 3,000 pounds 

per month while TCC produced a total of 400 pounds from intermittent operations over a few months. 

The modeled external dose is based on a 30-gallon drum containing 400 pounds of U3O8. The 

contract with the AEC specified a 30-gallon steel open head drum was to be used for delivery of the 

product to the AEC (AEC, 1953). 

Uranium Recovery Plant workers would have handled the uranium concentrate and been in close 

proximity to the material during certain steps in the process. Product would have been dried and then 

transferred to a drum, which may have been a manual process. Also, the material would have been 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

48 of 64 

sampled by TCC prior to shipment. Some other work in the plant may have required close work with 

the concentrate. Filtration of product may have required some direct hands-on work with the 

concentrated uranium product. To allow for the activities that may have required workers to be very 

close to bulk uranium concentrate, the external photon doses have been modeled based on work close 

to a drum that contained the entire 400 pounds of uranium produced by TCC. Although workers 

would not have been exposed to the full drum during the entire period, that scenario is assumed for the 

purpose of modeling a bounding external dose rate from exposure to uranium in the Uranium 

Recovery Plant. 

The external doses were modeled based on the assumption that the concentrate had the isotopic ratios 

listed in Table 5-3, which assumes dose contribution from natural uranium and progeny and 

contaminants in the natural thorium decay series. 

MCNPX (version 2.7b) was used to determine the dose rate per curie of 238U regardless of the actual 

activity in the drum. This was later adjusted for actual source activity and all radionuclides were 

considered as a ratio with respect to 238U to determine the number of photons and electrons per decay 

of 238U. The number and energy of photons per unit decay of 238U was compiled from ENSDF files 

from NUCDAT (11/28/1005) located at http://www.nndc.bnl.gov/nudat2/indx_dec.jsp. These 

emissions were binned by emission probability. The resulting total photon emission probability per 

decay of 238U was 1.0434 for the energy range of 0.001 to 3.1 MeV. The number and energy of beta 

particle emissions per unit decay of 238U was compiled from ENDSDF using the NUCDAT 11-28-

2005 database. These emissions were binned by emission probability. The resulting total beta 

emission probability per decay of 238U was 4.19. 

A density of 2 g/cm3 was chosen to provide a claimant-favorable geometry. Based on previous 

evaluations (for Blockson), the density is nominally a self-correcting factor for uranium in that it adds 

more shielding as it adds more activity. Since the total quantity was fixed at 400 lbs, the exposure 

geometry was allowed to be higher in the drum by using a lower density. The drum was assumed to 

have the same physical specifications as listed in ORNL No. 100-1A2-0006. This model results in a 

drum approximately 75% full. 

Photon flux was evaluated at 30 cm and 100 cm from the edge of the drum at a height of 77.9 cm 

above the floor, the approximate height of the testes. Factors from ICRP 74 were used to convert the 

photon flux to units of air kerma. Results are provided in Table 7-10. 

Table 7-10: Photons Dose Rates from Yellowcake

Distance from 

drum

Activity of 238U in 

drum (Ci)

Photon emission 

dose (rad/hr)

Bremsstrahlung 

dose (rad/hr)

Total dose rate

(rad/hr)

30 cm 0.05134 1.89E-3 2.02E-4 2.09E-3

100 cm 0.05134 6.16E-4 6.64E-5 6.82E-4

The exposure time is based on the assumptions for internal dose specified in Section 7.2.2. During the 

period of operation of the Uranium Extraction Plant, it is assumed the intermittent uranium extraction 

work was a two-day-per-week process and that a worker spent 20% of the time over those two days at 

a distance of 30 cm (one foot) from a drum of product that had been packaged. If a 10-hour work day 

is assumed, then a worker would spend 4 hours per week at a distance of 30 cm from the drum. This 

exposure scenario also allows for work in close proximity to uranium concentrates in other locations 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

49 of 64 

in the building, such as working with filter presses or manually transferring the material. If a 50-hour 

workweek is assumed, the worker could have been working in the building for an additional 46 hours 

at a lower exposure rate doing tasks that did not involve direct handling of uranium concentrate. 

Exposure during that time is modeled by assuming a general area dose rate that is equivalent to the 

dose rate at 100 cm from the drum of material. Thus the hypothetical maximally exposed worker’s 

weekly external whole-body photon dose is bounded by assuming 4 hours exposure at 30 cm and 46 

hours exposure at 100 cm from the drum of material. Applying the modeled air kerma hourly rates to 

these assumed hours results in a weekly dose (air kerma) of 0.040 rad; this is equivalent to 0.0057 rad 

per calendar day. 

7.3.3 External Dose during the Development Period

TCC received leached-zone phosphate ores from the AEC for contracted research work. The initial

work was reported to involve hundreds of grams of samples. A drum of the ore was also received in 

early 1954 and the AEC provided three 25-pound samples of leached-zone material to TCC in 1955. 

The development work involved experimental methods to separate phosphate from undesirable 

species in the leached-zone ore, such as alumina and silica. Uranium analyses were made along with 

the analyses of the primary species of concern (Kopf, 1954; Cutter, 1954; Greenleaf, 1955). 

Workers involved with the leached-zone research were handling far less material than those workers 

involved with phosphate production at TCC. The leached-zone ore had similar radiological 

constituents as the phosphate rock; the development work also did not involve the extraction of 

radionuclides, other than for sample analysis. Therefore, dose to workers involved with leached-zone 

ore research would have been much lower than dose to workers involved with production of fertilizers 

with TENORM. Since dose to workers from fertilizer operations is also covered during the proposed 

SEC period, evaluation of the much lower dose to a few workers from intermittent development is not 

needed. 

7.3.4 Shallow Dose 

Electron radiation is considered as a source of shallow dose for workers at TCC. Shallow dose from 

exposure to uranium product in the Uranium Recovery Plant and shallow dose from phosphate 

materials processed at TCC are considered. 

Radionuclides that contribute to electron dose from the source term specified in Table 5-3 for uranium 

recovery operations include Pa-234m and Bi-210 in addition to short-lived Ra-226 progeny, primarily 

Bi-214 and Pb-214. Several of the other radionuclides contribute marginally to the external dose. All 

of those radionuclides would be present prior to the phosphogypsum filtration step where most of the 

Ra-226 and progeny and some other radionuclides are removed with the phosphogypsum. 

7.3.4.1 Shallow Dose from the Phosphate Plant

The beta-emitting radionuclides in the various materials and wastes would be distributed throughout 

the material in low concentrations roughly equivalent to concentration of U-238 in phosphate rock. 

Some dry phosphate products may have enhanced U-238 concentrations if the phosphoric acid used 

was not diverted over to the Uranium Recovery Plant, which is probable because uranium production 

was shut down during all but a few months of the AEC covered period. The most significant beta 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

50 of 64 

dose from the source term would be from Pa-234m, which emits a 2.29 MeV maximum-energy beta 

particle. 

The beta dose from phosphates is derived from reported beta dose rates from yellowcake. Based on 

Figure 2 of NRC Regulatory Guide 8.30, the dose rate at 30 cm (one foot) from the surface of aged 

yellowcake is between 1 and 2 mrem/hr. For this evaluation 2 mrem/hr is presumed. The yellowcake 

beta dose rate at 30 cm is for U3O8 (aged for Pa-234m to be in equilibrium with U-238). The beta 

dose rate was from Pa-234m, Th-234 and Th-231, Pa-234m is the primary contributor to the dose and

has the highest maximum energy (2.29 MeV). An estimate will be made of the number of 2.29 MeV 

beta particles that are generated per unit time on the surface of yellowcake and the number that are 

generated per unit time on the surface of various phosphate materials. For this evaluation, the number 

of decays on the surface of the materials will be determined by using a material depth equivalent to 

the maximum range of the 2.29 MeV beta, which is different for the various materials. The factors

used in the calculations are selected to provide a bounding favorable ratio of the flux of beta particles 

from phosphates versus yellowcake. The ratio will be used to derive phosphate beta dose from 

yellowcake beta dose. The factors used, and the rationale, follow. 

Table 5-4 lists U-238-specific activities in several typical phosphate fertilizers. Diammonium 

phosphate (DAP), (NH4)2HPO4, was the highest at 63 pCi/g. This concentration is assumed for TCC 

based on the uranium not having been recovered from the phosphates (prior to fertilizer production) 

for sale to the AEC. DAP has a density of about 1.6 g/cm3

. The rock would have a relatively high 

percentage of P2O5 (density = 2.39 g/cm3

) and would include calcium carbonate (density < 3 g/cm3) 

and small amounts of aluminum and iron oxides as well as small amounts of other impurities. The 

phosphogypsum waste would have significantly less U-238 and Pa-234m. Other common phosphate 

fertilizers may have been produced; TCC produced dicalcium phosphate as animal feed. The 

chemical processes at TCC included handling large volumes of phosphates in a slurry from which the 

insoluble gypsum was filtered out of solution. 

The number of Pa-234m beta particles produced per cm2

 of the surface of the phosphate materials was 

calculated using assumptions selected to overestimate the true rate. Water was assumed to be the 

medium to overestimate the effective range of the 2.29 MeV beta particle in the phosphate products 

and slurries. The maximum range of a 2.29 MeV beta is approximately 1 cm in water, thus for 

estimating the number of Pa-234m betas per cm

2

, only the top 1 cm of material was considered. The 

U-238 and Pa-234m concentration are assumed to be 63 pCi/g. To bound the number of atoms in 1 

cm of material, an upper average density of 3 g/cm3 is assumed, based on a review of the densities of 

the various compounds. Assuming equilibrium with U-238, there are 0.98 beta particles (for the Pa234m 2.29 MeV max beta) for every decay of U-238, resulting in 411 betas produced per minute per 

cm2 of material (top 1 cm). This rate will be compared to the rate from the more highly radioactive 

surface of yellowcake, for which beta dose rates are available.

The density of yellowcake is assumed to be 2.055 g/cm3

. The maximum range of a 2.29 MeV beta 

particle in yellowcake would be less than the 1 cm range in water. The maximum range of a 2.29 

MeV beta particle in lead is approximately 0.1 cm. This is considered a reasonable value to use for 

this calculation to underestimate the maximum range in yellowcake. Thus, only the number of 2.29 

MeV betas from the top 0.1 cm of the surface of yellowcake is considered. Assuming equilibrium 

with U-238, there are 0.98 beta particles (for the Pa-234m 2.29 MeV max beta) for every decay of U238. Yellowcake is assumed to be 85% U-238. A U-238 specific activity of 7.5 x 105 dpm/g is used. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

51 of 64 

This results in the generation of 1.284 x 105 beta particles per minute per cm

2 of yellowcake (top 0.1 

cm). 

The calculated number of Pa-234m beta particles per cm

2

 from yellowcake was divided by the 

calculated number from the phosphate material. The number from yellowcake was higher by a factor 

of 312 (1.284 x 105 min-1/411 min-1). For comparison purposes, the concentration of the U-238 in 

yellowcake is several thousand times higher than it is in phosphate rock. The factor of 312 was 

determined by using favorable values to intentionally minimize the ratio for the purpose of deriving a 

reasonable bounding beta dose rate from the phosphate material. A check of the maximum range of 

beta particles was also made for a 0.5 MeV beta particle to see if the above ratio would be similar for 

beta particles of less energy. The difference in maximum range between a water absorber to a lead 

absorber is approximately a factor of 10, which is similar to the relative difference noted above for a 

2.29 MeV particle. Therefore, the estimated ratio is assumed to be favorable for a range of energies 

that contribute significantly to electron dose from phosphate materials.

The dose rate taken from the NRC publication is for aged yellowcake that accounts for dose from 

Pa-234m, Th-234, and Th-231. However, the source term in the phosphate plant at TCC is considered 

to have additional radionuclides according to the ratios presented in Section 5.3.2, Table 5-6. The 

skin dose coefficients for each radionuclide in Table 5-6, and associated progeny, were evaluated 

using Federal Guidance Report No. 12, Table III.1, Dose Coefficients for Air Submersion (EPA, 

1993). Those coefficients indicate that Pa-234m, Th-234, and Th-231 would account for about 23% 

of the skin dose for all the radionuclides in the phosphate source term at the relative concentrations 

assumed for phosphates in Table 5-6. 

The 2 mrem/hr yellowcake beta dose rate was multiplied by a factor of 4.3 (1/0.23) to allow for 

radionuclides in the phosphate source term that are not in aged yellowcake. Then the rate was divided 

by 312 to adjust for lower flux of beta particles from the surface of phosphates (relative to 

yellowcake), resulting in a dose equivalent rate of 0.028 mrem/hr. 

2 mrem/hr * 4.3 * 1/312 = 0.028 mrem/hr 

If a 2500-hour work year is assumed, the annual dose would be 70 mrem or 0.070 rem/year. Shallow 

dose is assigned as electrons with energy greater than 15 keV. 

7.3.4.2 Shallow Dose from the Uranium Recovery Plant

Beta dose to the skin is estimated by assuming work on uranium recovery 2 days per week from 

October 5, 1953, through March 31, 1954. Exposure to an open drum of uranium product or 

concentrated material in process would provide a source for shallow dose from electrons greater than 

that received from working with phosphate fertilizers. Using the assumptions in Section 7.3.2, it is 

assumed the intermittent uranium extraction work was a two-day per week process and that a worker 

spent 2 hours per day at a distance of 30 cm (one foot) from a concentrated uranium source and 8 

hours per day at a distance of 100 cm from the uranium concentrates. Beta dose in the Uranium 

Recovery Building is assumed insignificant for the remaining hours per week. However, if 50 hours 

worked per week is assumed, a worker could have received beta dose during an additional 30 hours 

per week from working with phosphates. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

52 of 64 

For the purpose of this evaluation, the uranium concentrate is assumed to be yellowcake to provide a 

favorable estimate of dose received at TCC. Based on Figure 2 of NRC Regulatory Guide 8.30, the 

dose rate at 30 cm (one foot) from the surface of aged yellowcake is between 1 and 2 mrem/hr. For 

this evaluation 2 mrem/hr is presumed. That rate is also used for the 100-cm distance. Using the time 

and distance exposure model from Section 7.3.2, it is assumed that workers were exposed to 

yellowcake for 20 hours per week. 

The rate taken from the NRC publication is from aged yellowcake that accounts for dose from 

Pa-234m, Th-234, and Th-231. However, the source term in the Uranium Recovery Plant is 

considered to have additional radionuclides according to the ratios presented in Section 5.3.1, Table 5-

3. The skin dose coefficients for each radionuclide in Table 5-3, and associated progeny, were

evaluated using Federal Guidance Report No. 12, Table III.1, Dose Coefficients for Air Submersion

(EPA, 1993). Those coefficients indicate that Pa-234m, Th-234, and Th-231 would account for about 

57% of the skin dose for all the nuclides in the Uranium Recovery Plant source term at the relative 

concentrations listed in Table 5-6. 

The 2 mrem/hr yellowcake beta dose rate was multiplied by a factor of 1.8 (1/0.57) to allow for 

radionuclides in the Uranium Recovery Plant source term that are not in aged yellowcake, resulting in 

a dose equivalent rate of 3.6 mrem/hr (0.0036 rem/hr). 

2 mrem/hr * 1.8 = 3.6 mrem/hr 

For 20 hours per week exposure in the Uranium Recovery Plant, this results in 72 mrem per week beta 

dose. For the remaining 30 hours per week, beta exposure from phosphates at a rate of 0.028 mrem/hr 

is presumed (see Section 7.3.4.1 above), resulting in an additional 0.84 mrem. This results in weekly 

beta exposure of 73 mrem (0.073 rem/week), or 0.00020 rem per calendar day from October 5, 1953, 

through March 31, 1954. 

Shallow Dose from Deposition on Skin

It is also assumed that there was a potential for workers to receive a shallow dose from electrons due 

to skin contaminated with yellowcake. The amount of skin contamination can be calculated by using 

the measured deposition velocity of 4-μm particles to skin of 0.012 m/s (Andersson, 2002; Fogh, 

1999), assuming that the material was deposited on the skin for an entire 8-hour shift. Modeled dose 

from this method is negligible when compared to the shallow dose estimate from the exposure to 

drums of yellowcake (discussed above) and the estimated skin dose from contaminated clothing 

discussed below. 

Beta Dose from Contaminated Clothing

Skin dose from contamination transfer to the skin, and from contact with contaminated work clothing, 

was also considered. Mallinckrodt Chemical Company dose rate studies from contaminated clothing 

were evaluated and average dose rates from contaminated clothing at Mallinckrodt indicate a level of 

1.5 mrem/hr (AEC, 1958). The Mallinckrodt dose rate is used as a bounding condition for TCC 

because Mallinckrodt handled materials of similar radiological constituents, but in larger quantities 

and with a higher radioactive material content. During operation of the Uranium Recovery Plant it is 

assumed that the workers were exposed to that level for 10 hours per week, which is considered an 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

53 of 64 

upper-bound condition. This results in a dose to the skin of 15 mrem/week, or 0.0021 rem per 

calendar day. Electron dose from work clothing in the phosphate plant is considered insignificant, 

compared to the other estimated doses, due to the relatively low concentration of radionuclides. 

Doses are applied as electrons > 15keV.

7.3.4.3 Extremity Dose

Extremity dose is considered for work with the uranium concentrates in the fertilizer plant and in the 

Uranium Recovery Plant. Former Blockson workers stated that during filtering operations, their 

hands were directly exposed to filter cake containing uranium, sometimes without gloves (OCASTKBS-0002). For the purpose of this evaluation, it is assumed that similar exposures occurred at 

TCC. Doses from filtering operations have been estimated for the hands and forearms; maintenance 

personnel may also have had extremity dose. 

Surface dose rates on yellowcake have been reported to be about 203 mrad per hour (DOE, 2000). 

For extremity dose in the fertilizer plant, the 203 mrad per hour rate for yellowcake was multiplied by 

a factor of 4.3 to allow for other radionuclides and divided by 312 to allow for the lower concentration 

of uranium in phosphate products (see Section 7.3.4.1), resulting in a beta dose rate of 2.8 mrad per 

hour. 

For the Uranium Recovery Plant extremity exposures, an estimate was made of shallow dose to the 

hands and forearms based on direct contact with pure yellowcake; yellowcake concentration in the 

product delivered to the AEC was estimated to be 40% to 60%. A factor of 1.8 was applied to the 203 

mrad per hour dose rate for yellowcake to allow for additional radionuclides (see Section 7.3.4.2), 

resulting in a beta dose rate of 365 mrad per hour. For dose modeling, operation of the Uranium 

Recovery Plant was assumed to have occurred over a 2-day period (see Section 7.3.4.2). Small 

amounts of uranium were recovered, and the modeled 2-day period per batch allowed for overnight 

drying and packaging of recovered yellowcake (see Section 7.2.2). For extremity dose, the time of 

direct contact with yellowcake in the Uranium Recovery Plant is assumed to be 1 hour per week 

during the period in which the plant operated. This results in a dose to the skin of the hands and 

forearms of 365 mrad per week. 

To bound extremity dose, calculations are performed for two periods: (1) from the start of plant 

operations in October 1953 through April 1954 doses are assumed to have occurred from both 

fertilizer plant and uranium plant work as done in Section 7.3.4.2; and (2) from May 1, 1954, through 

the end of the covered period exposure is modeled only from the fertilizer plant because the uranium 

plant was shut down. 

Prior to April 1954, workers are assumed to have 1 hour of extremity dose per week from the uranium 

plant (365 mrem per hour) and 5 hours of extremity dose per week from the phosphate plant (2.8 

mrem per hour), resulting in a dose of 379 mrem per week, or 0.054 rem per calendar day.

Starting May 1, 1954, 1 hour per day, 6 days per week, extremity exposure at the fertilizer plant (2.8 

mrem per hour) is presumed, resulting in a dose of 17 mrem per week, or 0.0024 rem per calendar 

day.
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

54 of 64 

7.3.5 External Dose from Residual Contamination 

External photon dose from residual contamination is considered. The modeled external dose rates

from uranium recovery operations were based on a model source term that would no longer be present 

because the drum of yellowcake would be gone (delivered to AEC). Dose from residual 

contamination in the Blockson facility is used to bound dose from residual contamination at TCC. 

Blockson’s uranium plant was in production for over seven years with essentially the same source 

term; dose rates on surfaces in that facility should provide a bounding estimate for TCC during the 

residual period. Based information in Section 5.0 of the Blockson Technical Basis Document, the 

photon dose from residual contamination at TCC will be assumed to be 0.06 R/yr (OCAS-TKBS0002). For TCC the dose rate will be applied as a constant because the TCC uranium plant produced 

less than 1% of the total uranium produced at Blockson. This dose rate is applicable to the Uranium 

Recovery Plant starting April 1, 1954, and continuing throughout the residual contamination period. 

However, the photon dose from the fertilizer plant is higher and will be applied through the end of the 

AWE period in September 1955. 

Beta dose from residual contamination is considered insignificant once the yellowcake was removed 

from the facility. 

7.3.6 External Dose from X-ray Examinations 

There is no specific information available on occupational X-ray exposure that workers may have 

received during the covered period. The formal contract (that replaced the 1952 letter contract) 

between the AEC and TCC in 1953 had a provision requiring TCC “to conform to all minimum health 

and safety regulations and requirements of the Commission” (AEC, 1953). The AEC considered an 

annual chest X-ray to be standard procedure for phosphate plant workers engaged in uranium recovery 

work (AEC, 1953b). Since there are no available records to determine which workers may have been 

required by TCC to have X-ray examinations, an annual chest X-ray examination is presumed to have 

been administered to all workers during the AEC covered period of 1953 through 1955. Doses should 

be assigned according to the recommendations in ORAU-OTIB-006. 

7.3.7 External Dose Summary

This evaluation estimated bounding external doses for members of the proposed class. The estimated 

doses are summarized in Table 7-11. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

55 of 64 

Table 7-11: Summary of External Doses for Texas City Chemicals

Time Period Dose1 Radiation Type Dose quantity

October 5, 1953, through 

March 31, 1954

0.0057 rad/day

Photons

50% 30 - 250 keV

50% > 250 keV

Air kerma

0.010 rem/day Electrons >15 keV Shallow dose from 

yellowcake

0.0021 rem/day Electrons >15 keV Shallow dose from 

contaminated clothing

0.054 rem/day

(operators and maintenance

workers) 

Electrons >15 keV Shallow dose to skin of 

hands and forearms

April 1, 1954, through 

March 31, 1955

0.00060 rem/day

Photons

50% 30 - 250 keV

50% > 250 keV

Deep dose equivalent

0.00020 rem/day Electrons >15 keV Shallow dose from 

phosphates

0.0024 rem/day

(operators and maintenance 

workers)

Electrons >15 keV Shallow dose to skin of 

hands and forearms

April 1, 1955, through

the end of covered 

employment

0.00016 R/day

Photons

50% 30 - 250 keV

50% > 250 keV

Roentgen

1 External doses are normalized to dose per calendar day. All values are considered constants.

7.3.8 External Dose Reconstruction Feasibility Conclusion 

This evaluation concludes that external dose reconstruction from uranium extraction operations for 

members of the proposed class is feasible based on source term information. Favorable exposure time 

estimates can be applied to modeled external dose rates. For work outside of the Uranium Recovery 

Plant, external dose data from other phosphate plants provides a basis for estimating dose to workers 

in the phosphate plant during the evaluated period. 

7.4 Evaluation of Petition Basis for SEC-00088

The following assertions, made on behalf of petition SEC-00088 for the Texas City Chemicals, Inc., 

site, were evaluated. Information and affidavit statements provided by the petitioner are summarized 

in the italicized statements below; the comments that follow are from NIOSH.

Radiation monitoring records for members of the proposed class may have been lost, 

falsified, or destroyed. 

Information regarding monitoring from TCC is unavailable. 

NIOSH has been unable to obtain any radiation monitoring records for members of the 

proposed class. At this time, it is not known if radiation monitoring records ever existed. 

If they were generated, they appear to have been lost or destroyed. In the absence of 

exposure data for TCC, NIOSH has developed claimant-favorable methods to estimate 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

56 of 64 

worker doses from uranium recovery work at TCC. These methodologies are described 

in Sections 7.2 and 7.3. 

7.5 Summary of Feasibility Findings for Petition SEC-00088

This report evaluates the feasibility for completing dose reconstructions for employees at the TCC 

from October 5, 1953, through December 31, 1956. NIOSH found that the process descriptions and 

source term data available are sufficient to complete dose reconstructions for the proposed class of 

employees with the exception of radon exposures through September 1955. 

NIOSH finds that all doses can be reconstructed after September 1955. 

Table 7-12 summarizes the results of the feasibility findings at TCC for each exposure source. 

Table 7-12: Summary of Feasibility Findings for SEC-00088

October 5, 1953 through December 31, 1956

Source of Exposure Reconstruction Feasible Reconstruction Not Feasible

Internal X

 - U-238 and associated progeny X

 - U-235 and associated progeny X

 - Th-232 and associated progeny X

 - Radon during AWE period

 (ended September 30, 1955) 

X

 - Radon during Residual Period X

External X

 - Gamma X

 - Beta X

 - Neutron X

 - Occupational Medical X-ray X

Although NIOSH found that it is not possible to completely reconstruct radiation doses for the 

proposed class, NIOSH intends to use any monitoring data that may become available for an 

individual claim (and that can be interpreted using existing NIOSH dose reconstruction processes or 

procedures). Therefore, dose reconstructions for individuals employed at Texas City Chemicals

during the period from October 5, 1953 through September 30, 1960, but who do not qualify for 

inclusion in the SEC, may be performed using these data as appropriate. 

8.0 Evaluation of Health Endangerment for Petition SEC-00088

The health endangerment determination for the class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(3). Under these requirements, if it is not 

feasible to estimate with sufficient accuracy radiation doses for members of the class, NIOSH must 

also determine that there is a reasonable likelihood that such radiation doses may have endangered the 

health of members of the class. Section 83.13 requires NIOSH to assume that any duration of 

unprotected exposure may have endangered the health of members of a class when it has been 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

57 of 64 

established that the class may have been exposed to radiation during a discrete incident likely to have 

involved levels of exposure similarly high to those occurring during nuclear criticality incidents. If 

the occurrence of such an exceptionally high-level exposure has not been established, then NIOSH is 

required to specify that health was endangered for those workers who were employed for a number of 

work days aggregating at least 250 work days within the parameters established for the class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC. 

Due to the lack of radon monitoring data for workers who were processing Ra-226-bearing materials 

in the production of phosphoric acid, and the lack of sufficient facility information to model potential 

radon exposures, it is not feasible to estimate radon exposures in the phosphoric acid plant at Texas 

City Chemicals with sufficient accuracy. There are no data to associate specific individuals with work 

in that plant or to conclude a worker was not exposed in the plant; therefore, NIOSH cannot estimate 

the maximum radiation dose for members of the NIOSH-proposed class with sufficient accuracy. 

NIOSH’s evaluation determined that it is not feasible to estimate radiation dose for members of the 

NIOSH-proposed class with sufficient accuracy based on the sum of information available from 

accessible resources. Modification of the class definition regarding health endangerment and 

minimum required employment periods, therefore, is required. 

9.0 Class Conclusion for Petition SEC-00088

The petitioner requested evaluation of all laborers from January 1, 1952, through December 31, 1956. 

Since the filing of the petition, the start date for TCC as an AWE facility was changed from January 1, 

1952, to October 5, 1953, and the end date was changed from December 31, 1956, to September 1955. 

In this evaluation, NIOSH considered exposures to all workers in order to bound the estimates of 

potential exposure for the petitioner’s proposed class of “laborers.” The petitioner-requested class 

was modified because radiation monitoring records are unavailable for TCC workers for the specified 

period, and all TCC employees were potentially exposed to radioactive materials as a result of 

AEC-related uranium extraction processes. 

Based on its research, NIOSH modified the petitioner-requested class to define a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes all Atomic Weapons Employer employees who worked at Texas City 

Chemicals, Inc., from October 5, 1953 through September 30, 1955 for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in combination 

with work days within the parameters established for one or more other classes of employees in the 

Special Exposure Cohort. 

NIOSH concluded that maximum radiation dose from residual contamination starting October 1, 

1955, can be reconstructed. 

NIOSH has carefully reviewed all material submitted by the petitioner, including the specific 

assertions stated in the petition, and has responded herein (see Section 7.4). NIOSH has also reviewed 

available technical resources and many other references, including the Site Research Database, for 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

58 of 64 

information relevant to SEC-00088. In addition, NIOSH reviewed its NOCTS dose reconstruction 

database to identify EEOICPA-related dose reconstructions that might provide information relevant to 

the petition evaluation. 

These actions are based on existing, approved NIOSH processes used in dose reconstruction for 

claims under EEOICPA. NIOSH’s guiding principle in conducting these dose reconstructions is to 

ensure that the assumptions used are fair, consistent, and well-grounded in the best available science. 

Simultaneously, uncertainties in the science and data must be handled to the advantage, rather than to 

the detriment, of the petitioners. When adequate personal dose monitoring information is not 

available, or is very limited, NIOSH may use the highest reasonably possible radiation dose, based on 

reliable science, documented experience, and relevant data to determine the feasibility of 

reconstructing the dose of an SEC petition class. NIOSH concludes that it has complied with these 

standards of performance in determining the feasibility or infeasibility of reconstructing dose for the 

class under evaluation. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

59 of 64 

10.0 References 

42 C.F.R. pt. 81, Guidelines for Determining the Probability of Causation Under the Energy 

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

Register/Vol. 67, No. 85/Thursday, p. 22,296; May 2, 2002; SRDB Ref ID: 19391. 

42 C.F.R. pt. 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule; May 2, 2002; SRDB Ref ID: 

19392. 

42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as Members of the Special 

Exposure Cohort Under the Energy Employees Occupational Illness Compensation Program Act of 

2000; Final Rule; May 28, 2004; SRDB Ref ID: 22001. 

42 U.S.C. §§ 7384-7385 [EEOICPA], Energy Employees Occupational Illness Compensation 

Program Act of 2000, as amended; OCAS website (http://www.cdc.gov/niosh/ocas). 

AEC, 1952, Letter Contract No. AT(49-1)-616, to Texas City Chemicals, Inc., from the Division of 

Raw Materials; U.S. Atomic Energy Commission; February 14, 1952; SRDB Ref ID: 16646, pp. 37-

48. 

AEC, 1953, Contract No. AT(49-1)-647, to Texas City Chemicals, Inc., from the Division of Raw 

Materials; U.S. Atomic Energy Commission, May 12, 1953; SRDB Ref ID: 16646, pp. 49-67. 

AEC, 1955a, Monthly Report on Activities of Domestic Production Phosphates, Division of Raw 

Materials During December 1955; USAEC, Division of Raw Materials, Domestic Production 

Phosphates, Office of Domestic Procurement; December 1955; SRDB Ref ID: 22264. 

AEC, 1955b, Eighteenth Semiannual Report of the Atomic Energy Commission, July 1955; SRDB Ref 

ID 37935. 

AEC, 1958, Symposium on Occupational Health Experience and Practices in the Uranium Industry; 

U.S. Atomic Energy Commission; October 15-17, 1958; SRDB Ref ID: 7886.

AEC, c1963, Receipts of U3O8 from Phosphate and other Miscellaneous Domestic Sources, undated, 

SRDB Ref ID 40645. 

Affidavit, site and medical information from survivor of former TCC employee; received April 4, 

2007; OSA document id: 103975.

Andersson, 2002, Radiation Dose Implications of Airborne Contaminant Deposition to Humans, K. G.

Andersson, et al; Health Physics 82(2), pp. 226-232; 2002; SRDB Ref ID: 33203. 

Barr, James A. Jr., 1955a, Assistant Director, Domestic Production, to C. L. Karl, Area Manager, 

Fernald Area, January 13, 1955, Washington D. C. 

Barr, James A. Jr., et al., 1955b, Recovering URANIUM as By-Product in Phosphate Processing,

Rock Products, October 1955, SRBD Ref ID: 30132. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

60 of 64 

Battelle-TBD-6001, Site Profiles for Atomic Weapons Employers that Refined Uranium and Thorium, 

Rev. F0; December 13, 2006; SRDB Ref ID: 30673. 

Battelle-TBD-6001, App. BH, Site Profiles for Atomic Weapons Employers that Refined Uranium and 

Thorium, Appendix BH – International Minerals and Chemical Corporation, Rev. 0; July 16, 2007; 

SRDB Ref ID: 35365. 

Burnet, Jr., George, 1957, Ammonium Phosphates via Wet Process Acid, Agriculture and Food 

Chemistry 5(4):258-263, April 1957. 

Christofano, Emil, and Harris, William B., 1960, The Industrial Hygiene of Uranium Refining; 

Environmental Health (Official Publication for the American Academy of Occupational Medicine), 

Vol. 1, No. 5; November 1960; SRDB Ref ID: 15774. 

Coppinger, E. A., and Rohrman, C. A., 1959, Ionium (Thorium-230) for Radioisotope Preparation 

(Status Report); Programming Operation, Hanford Laboratories Operation; December 15, 1959; 

SRDB Ref ID: 30884. 

Corporate Profile, 1958, Corporate Profile..., Journal of Agricultural and Food Chemistry, 1958: 

6(11), pp. 866-867, November 1958. 

Cortese, Frank, 1954, “$7,000,000 Texas City Plant Only a Starter,” Houston Chronicle, March 29, 

1953. 

Court records, 1989, Various excerpts, briefs and opinions from the U. S. District Court for the 

Southern District of Texas, Galveston Division, and the U. S. Court of Appeals for the Fifth Circuit, 

No. 88-2860, Amoco Oil Company, Plaintiff-Appellant, vs. Borden, Inc., Defendant-Appellee, SRDB 

Ref IDs 75115, 75138, 75156, 75158, 75161, 75165, 75177, 75180, 75184, 75185, 75187, 75190, and 

75193. 

Cutter, Paul R., et al., 1954, Texas City Chemicals Inc., to Geo. G. Marvin, Director, Research and 

Development, Division of Raw Materials, U. S. Atomic Energy Commission, letter summarizing work 

on contract AT(49-6)910, May 20, 1954; SRDB Ref ID 40643. 

DOE, 1986, Elimination Report, Amoco Chemical Company (the Former Texas City Chemicals, Inc.), 

Texas City, Texas; U.S. Department of Energy, Formerly Utilized Sites Remedial Action Program 

(FUSRAP), 1986; SRDB Ref ID 16646 (pages 8 - 13 of 197). 

DOE, 2000, DOE Standard, Guide of Good Practices for Occupational Radiological Protection in 

Uranium Facilities; U. S. Department of Energy (DOE), Washington D.C., August 2000; SRDB Ref 

ID 4617. 

Dunham, M.D., Charles L, 1953, Chief, Medical Branch, Physical Examinations for Phosphate 

Company Personnel Engaged in Uranium Recovery, February 2, 1953; SRDB Ref ID: 9728. 

Eidson, Arthur F., and Damon, Edward G., 1984, Predicted Deposition Rates of Uranium Yellowcake 

Aerosols Sampled in Uranium Mills, Health Physics, Vol. 46, No. 1, January 1984; SRDB Ref ID 

32091. 

Elzerman, Alan W., 2007, Report for Contract 5-15816, Review of Blockson Chemical Co. uranium 

recovery process, May 31, 2007; SRBD Ref ID 32738. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

61 of 64 

EPA, 1978, Radiological Surveys of Idaho Phosphate Ore Processing – The Wet Process Plant, U. S. 

Environmental Protection Agency (EPA), Technical Note ORP/LV-78-1, Las Vegas, Nevada, April 

1978; SRDB Ref ID 30886. 

EPA, 1993, Federal Guidance Report No. 12: External Exposures to Radionuclides in Air, Water, and 

Soil; K. F. Eckerman; J. C. Ryman; U.S. Environmental Protection Agency; Washington, D.C.; 

September, 1993; SRDB Ref ID: 11997. 

ERDA, 1976, Summary of Available Data on Specified Division of Raw Materials Contractors, memo 

and summary report from R. D. Nininger, Asst. Director for Raw Materials, Energy Research and 

Development Administration, to M. B. Biles, Director, Division of Safety, Standards and Compliance, 

U.S. Energy Research and Development Administration, October 5, 1976; SRDB Ref ID 16441 

FIPR, 1995, Microbiology and Radiochemistry of Phosphogypsum, Florida Institute of Phosphate 

Research (FIPR), Publication No. 05-035-115, May 1995; SRDB Ref ID 30136. 

FIPR, 1998, Evaluation of Exposure to Technically Enhanced Naturally Occurring Radioactive 

Materials (TENORM) in the Phosphate Industry, Florida Institute of Phosphate Research, Publication 

No. 05-046-155, July 1998; SRDB Ref ID 18076. 

Florida Meeting, 1955, Florida Production Meeting on Recovery of Uranium from Commercial 

Phosphoric Acid, Florida meeting summary; Author unknown; April 25-26, 1955; SRDB Ref ID 

29518. 

Fogh, C. L. et al., 1999, Quantitative Measurement of Aerosol Deposition on Skin, Hair and Clothing 

for Dosimetric Assessment – Final Report, June 1999; SRDB Ref ID: 14018 

Fukuma, H. T., et al., 2000, Distribution of Natural Radionuclides During the Processing of 

Phosphate Rock from Itataia-Brazil for Production of Phosphoric Acid and Uranium Concentrate; 

Radiochimica Acta., Vol. 88, pp. 809-812; 2000; SRDB Ref ID 30474. 

Gäfvert, T., Holm, E. and Roos, P., 2001, Radionuclide fluxes at a plant manufacturing dicalcium 

phosphate for domestic animals, Journal of Environmental Radioactivity, 54(2001)61-73. 

Greek, B. F. et al., 1957, Uranium Recovery from Wet Process Phosphoric Acid, Industrial 

Engineering Chemistry, Vol. 49, SRDB Ref ID: 35734. 

Greenleaf, 1955, Greenleaf, Elmer F., Process Development, Division of Raw Materials, to George G. 

Marvin, Assistant Director, Process Development, Division of Raw Materials, U.S. Atomic Energy 

Commission Office Memorandum, Visit to Texas City Chemical Company on January 12-13, 1955;

SRDB Ref ID 40644. 

Guimond, Richard J., and Windham Samuel T., 1975, Radioactivity Distribution in Phosphate 

Products, By-Products, Effluents, and Wastes, August 1975; SRDB Ref ID 13366. 

Hickey E. E., G. A. Stoetzel, et al., 1993, Air Sampling in the Workplace, NUREG-1400, U.S. Nuclear 

Regulatory Commission, Washington, D.C., 1993. 

Hull, C. D., and Burnett, W. C., 1996, Radiochemistry of Florida Phosphogypsum; Journal of 

Environmental Radioactivity, Vol. 32 (3) pp. 213-238, 1996; SRDB Ref ID 32562. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

62 of 64 

ICRP, 1979, Limits for Intakes of Radionuclides by Workers, International Commission on 

Radiological Protection (ICRP) Publication No. 30, Pergamon Press, Oxford; SRDB Ref ID 29220. 

ICRP, 1994a, Human Respiratory Tract Model for Radiological Protection, International Commission 

on Radiological Protection, Publication No. 66; Annals of ICRP, Vol. 24 (1-3), 1994; SRDB Ref ID:

22732. 

ICRP, 1994b, Dose Coefficients for Intakes of Radionuclides by Workers, International Commission 

on Radiological Protection, Publication No. 68, Annals of ICRP Vol. 24 (4); SRDB Ref ID 22731. 

ICRP, 1996, Conversion Coefficients for Use in Radiological Protection Against External Radiation, 

International Commission on Radiological Protection, Publication No. 74; Annals of ICRP Vol. 26 

(3/4); SRDB Ref ID: 22730. 

Johnson, 1953, Johnson, Jesse C., Director, AEC Division of Raw Materials, to Strauss, Lewis L., 

AEC Chairman; November 18, 1953; SRDB Ref ID: 16646 (page 24 of 197). 

Journal, 1953, Production of uranium as by-product of phosphate fertilizer industry making steady 

progress, Journal of Agriculture and Food Chemistry, Vol. 1(4), May 1953. 

Kopf, E. H., 1954, Kopf, E. H., Vice President & General Manager, Texas City Chemicals, Inc., to 

Geo. G. Marvin, Director, Research and Development, AEC Division of Raw Materials, letter 

summarizing work on contract AT(49-6)910, February 24, 1954, SRBD Ref ID 40640. 

Lardinoye, M. H., et al., 1982, Unexpected Build-up in Wet-process Phosphoric Acid Plants, Health 

Physics 42(4):503-514, April 1982. 

Lopker, E. B., 1951, Lopker, E. B., Director of Engineering, Blockson Chemical Co., to Wimpfen, 

Sheldon, Raw Materials Division, AEC; July 31, 1951; SRDB Ref. ID 9558. 

Menzel, Ronald G., 1968, Uranium, Radium, and Thorium Content in Phosphate Rocks and Their 

Possible Radiation Hazard, Journal of Agriculture and Food Chemistry, Vol. 16(2):231-234, May 

1953. 

Mills et al., 1977, Radiation Exposures in the Florida Phosphate Industry, International Radiation 

Protection Association (IRPA), IRPA 4 proceedings, P3-78, Paris, France. 

NCRP, 1993, Radiation Protection in the Mineral Extraction Industry, NCRP Report No. 118, 

Bethesda, MD.

NIOSH, 2007a, NIOSH Worker Outreach Meeting for Texas City Chemicals, Inc. (Smith-Douglass), 

October 18, 2007, 2:00 P.M. session, meeting minutes, http://www.cdc.gov/niosh/ocas. 

NIOSH, 2007b, NIOSH Worker Outreach Meeting for Texas City Chemicals, Inc. (Smith-Douglass), 

October 18, 2007, 7:00 P.M. session, meeting minutes, http://www.cdc.gov/niosh/ocas. 

NIOSH, 2007c, NIOSH Dose Reconstruction Project Worker Outreach Meeting with Former Workers 

from the Blockson Chemical Company; meeting transcripts and minutes, January 24-25, 2007, 

http://www.cdc.gov/niosh/ocas. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

63 of 64 

NRC, 2002a, Health Physics Surveys in Uranium Recovery Facilities, Regulatory Guide 8.30, Rev. 1, 

U.S. Nuclear Regulatory Commission (NRC), May 2002; SRDB Ref ID 13982. 

NRC, 2002b, Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario 

Using the Probabilistic RESRAD-Build 3.0 Code, NUREG/CR-6755, 

OCAS-IG-001, External Dose Reconstruction Implementation Guideline, Rev. 1, Office of

Compensation Analysis and Support; Cincinnati, Ohio, August, 2002; SRDB Ref ID: 22401. 

OCAS-IG-003, Radiation Exposures Covered for Dose Reconstructions under Part B of the Energy 

Employees Occupational Illness Compensation Program Act, Rev. 0, November 5, 2007. 

OCAS-IG-004, The Use of Data from Other Facilities in the Completion of Dose Reconstruction 

Under the Energy Employees Occupational Illness Compensation Program Act, Rev. 0; August 21, 

2008. 

OCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 0,

September 23, 2004; SRDB Ref ID 32022. 

OCAS-TIB-009, Estimation of Ingestion Intakes, Rev. 0, April 13, 2004; SRDB Ref ID: 22397. 

OCAS-TKBS-0002, Technical Basis Document for Atomic Energy Operations at Blockson Chemical, 

Joliet, Illinois, Rev. 02, National Institute for Occupational Safety and Health (NIOSH); Cincinnati,

Ohio; November 21, 2007; SRDB Ref ID: 36611. 

ORAUT-OTIB-0006, Dose Reconstruction from Occupationally Related Diagnostic X-ray 

Procedures, Rev. 03 PC-1; December 21, 2005; SRDB Ref ID: 20220. 

ORAUT-OTIB-0024, Estimation of Neutron Dose Rates from Alpha-Neutron Reactions in Uranium 

and Thorium Compounds, Rev. 00; April 7, 2005; SRDB Ref ID: 19445. 

ORAUT-OTIB-0043, Characterization of Occupational Exposure to Radium and Radon Progeny 

During Recovery of Uranium from Phosphate Materials, Rev. 00; January 6, 2006; SRDB Ref ID: 

22596. 

ORAUT-OTIB-0070, Dose Reconstruction During Residual Radioactivity Periods at Atomic 

Weapons Employer Facilities, Rev. 00, March 10, 2008; SRDB Ref ID 41603. 

ORAUT-TKBS-0002, Basis for Development of an Exposure Matrix for Blockson Chemical 

Company, Joliet, Illinois; Period of Operation: March 1, 1951 through March 31, 1962, Rev. 01, 

June 29, 2004, SRDB Ref ID 19480. 

ORNL, 1980, Preliminary Survey of Texas City Chemicals, Inc. (Borden Chemical Division of 

Borden, Inc.), Texas City, Texas; Health and Safety Research Division; Oak Ridge National 

Laboratory; March 1980; SRDB Ref ID: 16646 (page 70 of 197) 

Owen, P. T., and Hyder, L. K., Technologically Enhanced Natural Radioactivity: An Overview,

Report ORNL/CF-80/265, Oak Ridge National Laboratory, Oak Ridge, TN, 1980. 

Personal Communication, 2002, [Name Redacted], Communication regarding uranium program; 

various dates throughout 2002; SRDB Ref. ID 31328. 
SEC-00088 Rev.1 10/18/2010 Texas City Chemicals, Inc.

64 of 64 

Personal Communication, 2007a, Personal Communication with Former TCC Employee; Telephone 

Interview by ORAU Team, October 2, 2007; SRDB Ref ID 35466.

Personal Communication, 2007b, Personal Communication with Survivor of Former TCC Employee; 

Telephone Interview by ORAU Team; October 2, 2007; SRDB Ref ID: 35465.

Petition Form B, SEC-00088, with supporting information, March 13, 2007; OSA Ref ID 102669. 

Powers, Herman G., 1979, Letter regarding Lack of Uranium Recovery Activities; Herman G. Powers; 

March 22, 1979; SRDB Ref ID 16646, page 126. 

Proof of Relationship, 2007, affidavit of marriage and employment from survivor of former TCC 

employee; April 17, 2007; OSA Ref ID 102839. 

Robinson, B. M., 1955, Florida Meeting on Recovery of Uranium from Phosphates, memorandum to 

files; May 2, 1955; SRDB Ref ID 29526. 

Roessler, C. E., et al., 1979, Uranium and Radium-226 in Florida Phosphate Materials, Health 

Physics, Vol. 37(3): 269-277, February 1979; SRDB Ref ID 13364. 

S. Cohen & Associates, 2005, First Supplemental Review of the Bethlehem Steel Site Profile, ORAUTTKBS-0001, Draft Rev. 2, September, McLean, VA. 

Stoltz, Emil Jr., 1958, Recovery of Uranium from Phosphate Ores, July 24, 1958; SRDB Ref ID 4026. 

Web Elements, 2007, http://www.webelements.com/webelements/compounds/text/U/O8U3-

1344598.html, Sept. 27, 2007. 

Wilkinson, Gerald E. G., 1976, Uranium Recovery from Wet Process Phosphoric Acid: History & 

Present Status, Gardinier, Inc., May 17, 1976; SRDB Ref ID 10897. 

Windham, S., Partridge, J., and Horton, T., 1976, Radiation dose estimates to phosphate industry 

personnel, Publication EPA-520/5-76-014, Montgomery, AL. 

Wunder, 1954, Wunder, G. W., National Lead of Ohio Fernald Plant Manager, to Karl, C. L., AEC 

Fernald Area Manager, Official Sampling of Uranium Phosphates from Concentrates, December 29, 

1954.