SEC Petition Evaluation Report
 

Petition SEC-00103 

Report Rev #: 0 Report Submittal Date: 11/14/2008 

Subject Expert(s): James Mahathy, Robert Morris, Will McCabe, Darin Hekkala, Bryce Rich, Leo 

Faust, Sam Chu, Mel Chew, James Alexander, Ray Clark, Ed Scalsky 

Site Expert(s): None 

Petition Administrative Summary 

Petition Under Evaluation 

Petition # Petition 

Type 

Petition 

Qualification Date 

DOE/AWE Facility Name 

SEC-00103 83.13 March 4, 2008 Savannah River Site 

Related Petition Summary Information 

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

SEC00114 83.13 Savannah River Site Merged 

Related Evaluation Report Information 

Report Title DOE/AWE Facility Name 

None 

ORAU Lead Technical Evaluator: James Mahathy ORAU Review Completed By: Dan Stempfley 

Peer Review Completed By: [Signature on file] 

Timothy Taulbee 

11/14/2008 

Date 

SEC Petition Evaluation Reviewed By: [Signature on file] 

J. W. Neton 

11/14/2008 

Date 

SEC Evaluation Approved By: [Signature on file] 

Larry Elliott 

11/14/2008 

Date 

Petitioner Class Definition 

Construction workers and all other workers in all locations at the Savannah River Site, Aiken, SC, from 01/01/1950 to 

present. 

Class Evaluated by NIOSH 

All construction workers who worked in any area at the Savannah River Site during the period January 1, 1950 through 

December 31, 2007. 

NIOSH-Proposed Class(es) to be Added to the SEC 

None. The thorium operations prior to 1960 remain reserved as the evaluation continues 
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Evaluation Report Summary: SEC-00103, SRS 

This evaluation report by the National Institute for Occupational Safety and Health (NIOSH) 

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. 

Petitioner-Requested Class Definition 

Petition SEC-00103, qualified on March 4, 2008, requested that NIOSH consider the following class: 

Construction workers and all other workers in all locations at the Savannah River Site, Aiken, SC, 

from 01/01/1950 to present 

Class Evaluated by NIOSH 

Based on its preliminary research, NIOSH modified the petitioner-requested class. NIOSH evaluated 

the following class: All construction workers who worked in any area at the Savannah River Site 

during the period January 1, 1950 through December 31, 2007. 

NIOSH-Proposed Class(es) to be Added to the SEC 

Based on its full research of the class under evaluation, NIOSH has obtained worker monitoring data, 

area and air monitoring data, source term information, and process information that allow dose 

reconstruction to be performed with sufficient accuracy. Based on its analysis of these available 

resources, NIOSH found no part of the class under evaluation for which it cannot bound radiation 

doses with sufficient accuracy. However, NIOSH has reserved the feasibility determination for 

thorium exposures from January 1, 1950 through December 31, 1959; NIOSH is continuing to 

evaluate the thorium bounding approach for this time period. 

Feasibility of Dose Reconstruction 

Per EEOICPA and 42 C.F.R. § 83.13(c)(1), NIOSH has established (with the exception of the portion 

of the proposed class where the decision is reserved) that it has 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 the site profile and additional resources is sufficient to document or 

estimate the maximum internal and external potential exposure to members of the proposed class 

under plausible circumstances during the specified period. 

Health Endangerment Determination 

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is not required 

because NIOSH has determined that it has sufficient information to estimate dose for the members of 

the proposed class. 

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Table of Contents 

Evaluation Report Summary: SEC-00103, SRS ..................................................................................... 3
 

1.0 Purpose and Scope.......................................................................................................................... 9 

2.0 Introduction .................................................................................................................................... 9 

3.0 SEC-00103 Savannah River Site Class Definitions ..................................................................... 10 

3.1 Petitioner-Requested Class Definition and Basis ................................................................ 10 

3.2 Class Evaluated by NIOSH ................................................................................................. 12
 

3.3 NIOSH-Proposed Class(es) to be Added to the SEC .......................................................... 12 

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class ........................................................... 12 

4.1 Site Profile Technical Basis Documents (TBDs) ................................................................ 12 

4.2 ORAU Technical Information Bulletins (OTIBs) and Procedures ..................................... 13 

4.3 Facility Employees and Experts .......................................................................................... 14 

4.4 Previous Dose Reconstructions ........................................................................................... 15 

4.5 NIOSH Site Research Database .......................................................................................... 15 

4.6 Other Technical Sources...................................................................................................... 16 

4.7 Documentation and/or Affidavits Provided by Petitioners ................................................. 16 

5.0 Radiological Operations Relevant to the Class Evaluated by NIOSH ......................................... 18 

5.1 Savannah River Site Plant and Process Descriptions .......................................................... 18 

5.1.1 K, L, P, C and R Reactors and Associated Facilities .............................................. 19 

5.1.2 Heavy Water Rework Facility (400-D Area) .......................................................... 20 

5.1.3 Defense Waste Processing Facility (DWPF)........................................................... 21 

5.1.4 Saltstone Facility ..................................................................................................... 21
 

5.1.5 Tritium Facilities ..................................................................................................... 21
 

5.1.6 300 M-Area Fabrication Facilities........................................................................... 22 

5.1.7 Receiving Basin for Offsite Fuel (RBOF) and L-Basin .......................................... 23 

5.1.8 F Canyon ................................................................................................................. 23 

5.1.9 FB Line .................................................................................................................... 24 

5.1.10 JB Line..................................................................................................................... 24 

5.1.11 H Canyon ................................................................................................................. 24 

5.1.12 Old HB Line ............................................................................................................ 25 

5.1.13 New HB Line........................................................................................................... 26
 

5.1.14 Naval Fuels Manufacturing Facility........................................................................ 26 

5.1.15 Building 235-F Plutonium Fuel Fabrication Facility (PUFF) ................................. 27 

5.1.16 Savannah River Technology Center........................................................................ 27 

5.2 Radiological Exposure Sources from Savannah River Site Operations .............................. 27 

5.2.1 Internal Radiological Exposure Sources from SRS Operations .............................. 27 

5.2.2 External Radiological Exposure Sources from SRS Operations ............................. 28 

5.2.3 Incidents .................................................................................................................. 29 

6.0 Pedigree of Savannah River Site Data.......................................................................................... 29 

6.1 Internal Monitoring Data Pedigree Review......................................................................... 30 

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6.2 External Monitoring Data Pedigree Review........................................................................ 31 

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by NIOSH ...................................... 38 

7.1 Evaluation of Bounding Internal Radiation Doses at Savannah River Site ........................ 39 

7.1.1 Radionuclides of Internal Concern.......................................................................... 43 

7.1.1.1 Tritium...................................................................................................... 43 

7.1.1.2 Uranium.................................................................................................... 45 

7.1.1.3 Plutonium ................................................................................................. 46 

7.1.1.4 Neptunium (Np-237) ................................................................................ 48 

7.1.1.5 Americium (Am-241 and Am-243) .......................................................... 48 

7.1.1.6 Curium (Cm-244) ..................................................................................... 50 

7.1.1.7 Californium (Cf-252, including Cm-248 daughter) ................................. 51 

7.1.1.8 Thorium (Th-228, Th-232, and U-233) .................................................... 51 

7.1.1.9 Fission and Activation Products ............................................................... 54 

7.1.1.10 Cobalt-60 (Co-60) .................................................................................... 56 

7.1.1.11 Polonium................................................................................................... 56 

7.1.2 Internal Dose Reconstruction Feasibility Conclusion ............................................. 57 

7.2 External Radiological Exposure .......................................................................................... 57 

7.2.1 Photon Exposures .................................................................................................... 58 

7.2.1.1 Source Term ............................................................................................. 58 

7.2.1.2 Personnel Monitoring Data....................................................................... 59 

7.2.1.3 Feasibility of Estimating or Bounding Photon Dose ................................ 61 

7.2.2 Neutron Exposures .................................................................................................. 61 

7.2.2.1 Source Term ............................................................................................. 62 

7.2.2.2 Neutron Personnel Monitoring Data ........................................................ 64 

7.2.2.3 Feasibility of Bounding Neutron Doses ................................................... 66 

7.2.3 External Dose Reconstruction Feasibility Conclusion ............................................ 67 

7.3 Evaluation of Petition Basis for SEC-00103 ....................................................................... 68 

7.3.1 Dose Reconstructions for Construction Workers .................................................... 68 

7.3.2 Consideration of Construction Workers’ Involvement in Incidents........................ 68 

7.3.3 Exposure Patterns of Construction Workers ........................................................... 68 

7.3.4 1990 Tiger Team Assessment ................................................................................. 69 

7.3.5 Work in Non-radiological Areas ............................................................................. 70 

7.3.6 Cover-up of Incidents .............................................................................................. 71 

7.4 Other Potential SEC Issues Relevant to the Petition Identified During the Evaluation ...... 71 

7.4.1 Radionuclides from Special Campaigns Not Included in the TBD......................... 71 

7.4.2 Tank Farm Exposures .............................................................................................. 71 

7.4.3 Forestry Workers ..................................................................................................... 72 

7.4.4 Special Tritium Compounds .................................................................................... 72 

7.4.5 Bounding of Doses from Intake of Thorium ........................................................... 72 

7.4.6 Completeness of the HPAREH Database ................................................................ 73 

7.4.7 Dosimetry Uncertainty ............................................................................................ 74 

7.4.8 Early Construction Worker Monitoring Data .......................................................... 77 

7.4.9 Early Construction Worker Neutron Monitoring Data............................................ 77 

7.4.10 Neutron to Photon Ratio .......................................................................................... 77 

7.5 Summary of Feasibility Findings for Petition SEC-00103.................................................. 78 

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8.0 Evaluation of Health Endangerment for Petition SEC-00103...................................................... 78 

9.0 Class Conclusion for Petition SEC-00103 ................................................................................... 79 

10.0 References .................................................................................................................................... 80 

Attachment 1: Savannah River Site Events Chronology....................................................................... 90 

Figures 

6-1: Comparison of the Number of Monitored Workers in HPAREH vs. WSRC-RP-95-234 and 

Hardcopy 4th Quarter Reports Obtained from SRS in August 2008 ............................................ 33 

6-2: Comparison of Annual Dose Distributions for Construction Trades Workers in HPAREH and 

the 4th Quarter Reports ................................................................................................................. 34 

6-3: Comparison of Annual Dose Distributions for Construction Trades Workers in HPAREH and 

the 4th Quarter Reports ................................................................................................................. 35 

6-4: Comparison of 1960 Annual Dose Distributions Using the Annual Roll, 4th Quarter Reports, and 

HPAREH ...................................................................................................................................... 36 

7-1: SRS Isotope Production vs. Internal Monitoring for Intakes ....................................................... 40 

7-2: Number of Beta/Gamma (Photon) Film Dosimeter Badges Processed per Cycle for the 200-F 

Area. ............................................................................................................................................. 65 

7-3: Number of Neutron Film Dosimeter Badges Processed per Cycle for the 200-F Area. .............. 65 

7-4: Directional Response of Savannah River Plant Film Badge (Gamma)........................................ 75 

7-5: SRS Film Dosimeter Audit Results for March 1959 through September 1963 ........................... 76 

Tables 

4-1: No. of Savannah River Claims Submitted Under the Dose Reconstruction Rule ........................ 15 

5-1: Potential Contributors to SRS Internal Dose ................................................................................ 28 

6-1: Summary of Available Monitoring Data for SRS Workers ......................................................... 32 

6-2: Comparison of Construction Trades Workers to HPAREH........................................................ 35 

6-3: Comparison of Bonner Sphere, TLND, and NTA Film Measurements at the SRS Plutonium
 

Finishing Area .............................................................................................................................. 37 

7-1: Bioassay Start Dates for Specific Isotopes and Suitable Bioassay Technique............................. 41 

7-2: SRS Tritium Exposure Experience............................................................................................... 45 

7-3: Sources of Neutrons ..................................................................................................................... 62 

7-4: Typical Neutron Dose Rates on Special Targets .......................................................................... 63 

7-5: Numerical Summary of Tiger Team Radiation Protections Findings .......................................... 69 

7-6: Summary of Feasibility Findings for SEC-00103 ........................................................................ 78 

A-1: Savannah River Site Events Chronology ..................................................................................... 90 

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SEC Petition Evaluation Report for SEC-00103
 

1.0 Purpose and Scope 

This report evaluates the feasibility of reconstructing doses for all construction workers who worked 

in any area at the SRS during the period January 1, 1950 through December 31, 2007. 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 Office of Compensation Analysis and Support’s (OCAS) Internal 

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

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.1 

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 with sufficient accuracy radiation doses 

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

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 

1

 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. 

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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.2 

3.0 SEC-00103 Savannah River Site Class Definitions 

The following subsections address the evolution of the class definition for SEC-00103, Savannah 

River Site (SRS). When a petition is submitted by a claimant, the requested class definition is 

evaluated as submitted. If the available site information and data justify a change in the petitioner’s 

class definition, NIOSH will specify a modified class to be fully evaluated. After a complete analysis, 

NIOSH will determine whether to propose a class for addition to the SEC and will specify that 

proposed class definition. 

3.1 Petitioner-Requested Class Definition and Basis 

Petition SEC-00103, qualified on March 4, 2008, requested that NIOSH consider the following class 

for addition to the SEC: Construction workers and all other workers in all locations at the Savannah 

River Site, Aiken, SC, from 01/01/1950 to present. 

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

accurate dose reconstruction over time is impossible for the Savannah River Site workers in question 

based on “radiation exposures and radiation doses potentially incurred by members of the proposed 

class were not monitored either through personal monitoring or through area monitoring.” The 

SEC00103 petitioners asserted the following in the SEC petition: 

“Since the inception of the EEOICP [sic], the building trades have asked NIOSH to come up with 

a unique approach to construction worker dose reconstructions that will take into account the 

unique employment patterns and unreliable dose monitoring. To date, NIOSH has failed to do so. 

In 2005, a study was performed by the Center to Protect Workers’ Rights which has been provided 

to NIOSH compared 2,335 construction workers, who had been employed at the SRS site and who 

2 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. 

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had participated in the Former Worker Medical Screening Program for the SRS site, to the 

radiation dose records data set for the SRS site (known as HPAREH). A significant number of 

SRS construction workers have either no deep dose or all recorded “zero” doses in HPAREH. 

Based on HPAREH database of radiation monitoring records from SRS, it appears that underlying 

dose data are deficient for 50-90% of the construction workers employed at SRS. NIOSH has not 

explained how it can complete dose reconstructions in light of this deficiency. On May 10, 2003, 

NIOSH issued a site profile document for the SRS site which aimed to provide methods for dose 

reconstruction where individual worker monitoring records were deficient. Construction workers 

who had extensive employment experience from all phases of the SRS site operation met with 

NIOSH in Augusta on November 11, and identified deficiencies in the site profile document as it 

related to construction workers in a number of areas. Their concerns were also presented to the 

NIOSH Board on Radiation and Worker Health [sic] on December 9, 2003, to make sure there is a 

record of them at NIOSH. The opening comments from the Building Trades November 11th 

meeting are attached, as are comments made to the Advisory Board on December 9, 2003. The 

SRS site profile was revised a number of times in 2004 and 2005, but none of these modifications 

included the concerns raised by the building trades. 

There is no recent evidence to suggest that there is any reason to have confidence in the dose 

reconstructions performed by NIOSH. In a Congressional hearing on October 23, Mr. Shelby 

Hallmark of the Department of Labor testified that in 2007 DOL had returned 2,811 dose 

reconstruction cases for re-work, due to deficiencies identified in the work that NIOSH had 

performed. After re-working these cases, 385 cases which had been denied were approved. In 

other words, 14% had been wrong the first time around. Further, Mr. Hallmark stated that DOL 

would soon send another 4,400 cases back to NISOH [sic], and in addition to that 5,000 more. 

This means that DOL will have sent back half of the dose reconstruction cases completed. 

We conclude that in the six years that have elapsed since this program was implemented, NIOSH 

does not have a valid method to perform dose reconstructions for construction workers, and has 

not acted to rectify the deficiencies identified in the underlying knowledge base for the SRS site. 

Therefore, we believe that dose reconstructions on SRS construction workers cannot be performed 

with the reliability intended by the Act, and therefore, the construction workers employed at the 

SRS site should be included in the SEC.” 

NIOSH has concluded that there is sufficient information and documentation, and a defined dose 

reconstruction method (included in the SRS site profile document, ORAUT-TKBS-0003), for all other 

SRS non-construction workers. The dose reconstruction approach in ORAUT-TKBS-0003 for all SRS 

non-construction workers, coupled with the available personnel monitoring data, supports NIOSH’s 

ability to bound the dose for all SRS non-construction workers. Based on this information, NIOSH 

finds that there is insufficient support for the petition basis for SRS non-construction workers. 

In addition, based on its SRS research and data capture efforts, NIOSH determined that it has access 

to worker, co-worker, area and air radiological monitoring and source term data for SRS construction 

trade workers during the time period under evaluation. However, NIOSH considered the information 

and statements provided by the petitioner sufficient to qualify the petition for further consideration by 

NIOSH, the Board, and HHS. The details of the petition basis are addressed in Section 7.4. 

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3.2 Class Evaluated by NIOSH 

Based on its preliminary research, NIOSH modified the petitioner-proposed class in accordance with a 

ruling by the HHS administrative review panel (Branche, 2008). Therefore, NIOSH defined the 

following class for further evaluation: All construction workers who worked in any area at the 

Savannah River Site during the period January 1, 1950 through December 31, 2007. 

3.3 NIOSH-Proposed Class(es) to be Added to the SEC 

Based on its research, NIOSH has obtained monitoring data, source term information, and process 

information that allow dose reconstruction to be performed with sufficient accuracy. Based on its 

analysis of these available resources, NIOSH found no part of the class under evaluation for which it 

cannot estimate radiation doses with sufficient accuracy. However, NIOSH has reserved the 

feasibility determination for thorium exposures from January 1, 1950 through December 31, 1959; 

NIOSH is continuing to evaluate the thorium bounding approach for this time period. 

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class 

ATTRIBUTION: Section 4.0 and its related subsections were completed by Mike Mahathy, Oak Ridge 

Associated Universities. These 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. 

NIOSH identified and reviewed numerous data sources to determine information relevant to 

determining the feasibility of dose reconstruction for the class of employees under evaluation. This 

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

industrial processes, and radiation source materials. The following subsections summarize the data 

sources identified and reviewed by NIOSH. 

4.1 Site Profile Technical Basis Documents (TBDs) 

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. The Site Profile for Savannah River site is a single document with 

sections 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. 

As part of NIOSH’s evaluation detailed herein, it examined the following TBDs for insights into SRS 

operations or related topics/operations at other sites: 

•	 TBD: Savannah River Site, ORAUT-TKBS-0003; Rev. 02; April 5, 2005; SRDB Ref ID: 20176 

•	 Savannah River Site TBD Revisions, OCAS-PER-030; Rev. 0; December 12, 2007; SRDB Ref ID: 

38872 

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4.2 ORAU Technical Information Bulletins (OTIBs) and Procedures 

An ORAU Technical Information Bulletin (OTIB) is a general working document that provides 

guidance for preparing dose reconstructions at particular sites or categories of sites. An ORAU 

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 OTIBs and procedures as part of its evaluation: 

•	 OTIB: Maximum Internal Dose Estimates for Savannah River Site Claims, ORAUT-OTIB-0001, 

Rev 00; July 15, 2003; SRDB Ref ID: 19407 

•	 OTIB: Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures, ORAUTOTIB-0006, Rev. 03 PC-1; December 21, 2005; SRDB Ref ID: 20220 

•	 OTIB: Assignment of Environmental Internal Doses for Employees Not Exposed to Airborne 

Radionuclides in the Workplace, ORAUT-OTIB-0014, Rev 00; June 22, 2004; SRDB Ref ID: 

19432 

•	 OTIB: Interpretation of Dosimetry Data for Assignment of Shallow Dose, ORAUT-OTIB-0017, 

Rev. 01; October 11, 2005; SRDB Ref ID: 19434 

•	 OTIB: Internal Dose Overestimate for Facilities with Air Sampling Programs, ORAUT-OTIB0018,, Rev. 01; August 9, 2005; SRDB Ref ID: 19436 

•	 OTIB: Use of Coworker Dosimetry Data for External Dose Assignment, ORAUT-OTIB-0020, 

Rev. 01; October 7, 2005; SRDB Ref ID: 19440 

•	 OTIB: External Coworker Dosimetry Data for the Savannah River Site, ORAUT-OTIB-0032; 

November 7, 2006; SRDB Ref ID: 29964
 

•	 OTIB: Parameters to Consider When Processing Claims for Construction Trade Workers, 

ORAUT-OTIB-0052; Rev.00 PC-1, January 16, 2007; SRDB Ref ID: 29978
 

•	 OTIB: Internal Dose Reconstruction, ORAUT-OTIB-0060, Rev 00, February 6, 2007, SRDB Ref 

ID: 29984
 

•	 OTIB: Occupational X-Ray Dose Reconstruction for DOE Sites, ORAUT-PROC-0061, Rev. 01; 

July 21, 2006; SRDB Ref ID: 29987 

•	 OTIB: Calculation of Dose from Intakes of Special Tritium Compounds, ORAUT-OTIB-0066, 

Rev. 00; April 26, 2007; SRDB Ref ID: 31421 

•	 OTIB: Use of Claimant Data Set for Coworker Modeling, ORAUT-OTIB-0075, Rev. 00; no date 

yet; SRDB Ref ID: To be published in 2008 

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4.3 Facility Employees and Experts 

NIOSH interviewed SRS employees in order to obtain input regarding health physics practices, 

internal and external dosimetry programs, dose recording practices, and radiological incidents. 

•	 Personal Communication, 2008a, Personal Communication with Current Worker, Summary of 

Discussion of Selected Cohen & Associates SRS Site Profile Comments; Telephone Interview by 

ORAU Team; August 14, 2006; SRDB Ref ID: 27056 

•	 Personal Communication, 2008b, Personal Communication with Former Worker, Use of NTA 

Neutron Personnel Dosimetry in 1955-1968 at SRS; Telephone Interview by ORAU Team; March 

9, 2008; SRDB Ref ID: 48459 

•	 Personal Communication, 2008c, Personal Communication with Current Worker, SRS Incident 

Database, Special Hazards Investigations Database, HPAREH Database and Intake Registry; 

Telephone Interview by ORAU Team; April 3, 2008; SRDB Ref ID: 45062 

•	 Personal Communication, 2008d, Personal Communication with Engineer Concerning Thorium, 

Telephone Interview by ORAU Team; August 12, 2008; SRDB Ref ID: Not yet released by DOE 

•	 Personal Communication, 2008e, Personal Communication with Metallurgical Lab Technician, 

Telephone Interview by ORAU Team; August 11, 2008; SRDB Ref ID: Not yet released by DOE 

•	 Personal Communication, 2008f, Personal Communication with Electrician, Telephone Interview 

ORAU Team; August 11, 2008; SRDB Ref ID: Not yet released by DOE 

•	 Personal Communication, 2008g, Personal Communication with Former Health Physics 

Manager, Telephone Interview ORAU Team; August 11, 2008; SRDB Ref ID: Not yet released 

by DOE 

•	 SEC Worker Outreach Meeting for the Savannah River Site, National Institute for Occupational 

Safety and Health; May 22, 2008, 1 PM; SRDB Ref ID: Released by DOE; not yet in SRDB 

•	 SEC Worker Outreach Meeting for the Savannah River Site, National Institute for Occupational 

Safety and Health; May 22, 2008, 6 PM; SRDB Ref ID: Not yet released by DOE 

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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. (NOCTS data available as of October 1, 2008) 

Table 4-1: No. of Savannah River Claims Submitted Under the Dose Reconstruction Rule 

Description Totals 

Total number of claims submitted for dose reconstruction 3264 

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

under evaluation (Construction / Building Trades from January 1, 1950 through December 31, 2007) 1798 

Number of dose reconstructions completed for energy employees who meet the definition criteria for 

the class under evaluation (i.e., the number of such claims completed by NIOSH and submitted to the 

Department of Labor for final approval). 1358 

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

evaluated class definition 1467 

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

evaluated class definition 1474 

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

records could be obtained for the employee. Dose reconstructions have been completed based on 

actual monitoring data (data reported for over 80% of construction worker claims) and with the use of 

ORAUT-OTIB-0052. 

4.5 NIOSH Site Research Database 

NIOSH also examined the Site Research Database (SRDB) to locate documents supporting the 

evaluation of the proposed class. As of March 2008, there were about 600 technical documents in the 

SRDB pertaining to the Savannah River Site. In order to address the petitioner’s concerns and issues, 

NIOSH undertook a much larger on-site data review from June through August 2008. As a result, 

approximately 500 additional documents have been captured and reviewed (or are in the process of 

being reviewed, as is the case for the pre-1960 thorium assessment). In addition to technical reports, 

NIOSH has collected: (1) all of the Quarterly External Dosimeter reports since 1958; (2) all of the 

bioassay logbooks which contain the individual urinalysis sample results; and (3) the Site Special 

Hazards Investigation Reports (Incident Reports). 

Documents evaluated for relevance to this petition include historical background on the evolution of 

the site, process descriptions, radiological monitoring data (surface and air concentrations, personnel 

external and internal exposures), information on the radiological controls program as well as monthly 

reports, incident documentation, and epidemiological studies. 

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4.6 Other Technical Sources 

•	 OCAS-TIB-006, Interpretation of External Dosimetry Records at the Savannah River Site (SRS), 

Rev. 2; National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; October 

4, 2007; SRDB Ref ID: 35409 

•	 OCAS-TIB-007, Neutron Exposures at the Savannah River Site, Rev. 01; National Institute for 

Occupational Safety and Health (NIOSH); Cincinnati, Ohio; October 15 2007; SRDB Ref ID: 

35675 

•	 Health Protection Annual Radiation Exposure History Database (HPAREH) 

•	 Health Protection Radiation Exposure Database (HPRED) 

•	 SRS Site Incident Database 

•	 Internal Dose Registry 

4.7 Documentation and/or Affidavits Provided by Petitioners 

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

petitioners: 

•	 Petition Attachment 1 – NIOSH/Union SPD Meeting, November 11, 2003, November 19,2007; 

OSA Ref ID: 104377 

- Includes a review of the SRS SPD by former SRS union Energy Employees. 

- Specifically discusses the review and assessment of the SRS SPD and makes a conclusion 

about the available personnel monitoring data. 

•	 Petition Attachment 2 – Statement from AFL-CIO Science Advisor to NIOSH and the Board, 

December 9, 2003, November 19, 2007; OSA Ref ID: 104377 

- CPWR statements made regarding SRS construction workers and its evaluation of the 

EEOICPA dose reconstruction program, conflict of interest, and Rule conflicts. 

•	 Petition Attachment 3 – CPWR-NIOSH Meeting on Variance in Construction Worker Radiation 

Exposure Monitoring, July 27, 2005, November 19, 2007; OSA Ref ID: 104377 

- CPWR statements made regarding SRS construction workers and its evaluation of the 

EEOICPA dose reconstruction program and dose reconstruction methodology. 

•	 Petition Attachment 4 – What is the best estimate of the daily ventilation rate for construction 

workers? July 28, 2005, November 19, 2007; OSA Ref ID: 104377 

- CPWR statements made regarding SRS construction workers and its evaluation of the 

EEOICPA dose reconstruction program and dose reconstruction methodology. The 

attachment includes a PowerPoint presentation and a newspaper article. 

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•	 Affidavit, “Non-specific claimed radiological situations by former Energy Employee,” November 

19, 2007; OSA Ref ID: 104377 

•	 Affidavit , “Non-specific claimed radiological situations by survivor,” November 19, 2007 OSA 

Ref ID: 104377 

•	 Affidavit, “Semi-specific claimed contamination incidents by former Energy Employee – no data 

and no way to confirm,” November 19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Semi-specific claimed contamination incidents by former Energy Employee,” 

November 19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Semi-specific claimed radiological situations by former Energy Employee,” November 

19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Non-specific claimed radiological situations by former Energy Employee,” November 

19, 2007; OSA Ref ID: 104377 

• Affidavit, “Non-specific claimed radiological situations and chemical exposures by former 

Energy Employee,” November 19, 2007 OSA Ref ID: 104377 

•	 Affidavit, “Non-specific claimed radiological situations and beryllium exposures by former 

Energy Employee,” November 19, 2007 OSA Ref ID: 104377 

•	 Affidavit, “Semi-specific claimed radiological situations (contaminated railroad cross-ties) by 

former Energy Employee,” November 19, 2007 OSA Ref ID: 104377 

•	 Affidavit, “Non-specific claimed radiological situations by former Energy Employee,” November 

19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Semi-specific claimed radiological situations by former Energy Employee,” November 

19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Non-specific claimed radiological situations associated with dosimeters by former 

Energy Employee,” November 19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Non-specific claimed radiological situations by former Energy Employee,” November 

19, 2007; OSA Ref ID: 104377 

•	 Affidavit, “Non-specific claimed radiological by a former Energy Employee,” November 19, 

2007; OSA Ref ID: 104377 

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5.0 Radiological Operations Relevant to the Class Evaluated by 

NIOSH 

The following subsections summarize both radiological operations at the SRS from January 1, 1950 

through December 31, 2007 and the information available to NIOSH to characterize particular 

processes and radioactive source materials. From available sources NIOSH has gathered process and 

source descriptions, information regarding the identity and quantities of each radionuclide of concern, 

and information describing both the processes through which radiation exposures to construction 

workers may have occurred and the physical environment in which they may have occurred. 

Construction workers were used in all of the processes and operations discussed in Section 5.1, 

although some work was new construction not involving the presence of radiological materials 

(DPSP-55-454-2). The information included within this evaluation report is intended only to be a 

summary of the available information. 

5.1 Savannah River Site Plant and Process Descriptions 

ATTRIBUTION: Section 5.1 and its related subsections were completed by James K. Alexander, Oak 

Ridge Associated Universities. These 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. 

In 1950, the Atomic Energy Commission (AEC) selected a site in southeast South Carolina for 

construction of nuclear reactor facilities and ancillary operations that would manufacture plutonium, 

tritium, and other materials needed for the assembly of thermonuclear weapons. It was roughly 

circular in shape, slightly over 20 miles in diameter, and covered approximately 375 square miles. 

The location was chosen because of its proximity to the Savannah River, a large water source needed 

to remove the heat generated by the reactors and as a source of relatively clean water for heavy water 

extraction. 

The five heavy-water-moderated production reactors subsequently built on the site (nominally 

designated as the C, L, P, K, and R Reactors), the F and H “canyon” chemical processing areas, and 

the D-Area Heavy Water Production Facilities were spaced two to three miles apart along a roughly 

seven-mile diameter circle placed near the geometric center of the site. This configuration was chosen 

to facilitate security and public safety considerations (Bebbington, 1990). 

Over the site’s entire operational history, more than 1,000 separate facilities have been established, 

concentrated on only 10 percent of the total land area (DOE, 2008). A chronology of SRS 

development and the various production facilities is provided in Attachment 1. 

The material production activities associated with routine operation of the SRS reactors began in 

December 1953 with the start-up of the R Reactor, and continued until 1964 when demand for 

plutonium and tritium had significantly decreased; this reactor was then shut down. These decreasing 

demand trends slowly continued and the C Reactor was shut down in 1985. The K and P reactors 

were shut down in 1988. Operation of the L Reactor continued sporadically until 1991, when it was 

also shut down. K Reactor was restarted briefly in 1992 to facilitate the addition of a new cooling 

tower, but it was placed on “standby” by 1993. By 1996, all five SRS production reactors were 

considered to be in permanent “cold shutdown” status; none have been operated since. Between 1953 

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and 1988, the SRS reactors produced about 36 metric tons of plutonium (SRS Highlights, 2000; 

Bebbington, 1990; DOE, 2008; DOE/DP-0137). 

Currently, the SRS supports various DOE environmental restoration missions associated with waste 

management, radioactive material vitrification, special nuclear material storage, research and 

development, and technology transfer. The only remaining weapons program mission at SRS is the 

operation of the Replacement Tritium Facility (RTF) that began in 1994 (SRS, 2008). Construction 

workers have worked at all of the facilities in construction and maintenance operations. 

5.1.1 K, L, P, C and R Reactors and Associated Facilities 

The five SRS production reactors were all heavy-water-moderated designs ranging in power levels 

from 2,400 MW-thermal to 3,000 MW-thermal. They produced plutonium and tritium for use in 

nuclear weapons and, from time to time, produced other isotopes for various purposes. 

The reactors were built in sequence so that the first-completed could be turned over for operations and 

started up while construction on the remaining (geographically separated) reactors continued. The 

reactor buildings were designed to prevent the release of direct radiation and/or significant quantities 

of radioactive materials. The reactor enclosures were designed to withstand possible enemy attack, 

which was compatible with advanced containment safety systems. 

The SRS reactors differed substantially from the “graphite pile” reactors that had been previously built 

and operated at Hanford in that heavy water was used as both a neutron moderator and as the primary 

reactor coolant. Water from the Savannah River was used as the secondary coolant for all five of the 

reactors. The SRS reactor vessels were all made of stainless steel, about 16 feet in diameter, with 

precisely located holes at the top and corresponding pins at the bottom that held the reactor core lattice 

of fuel rods, target assemblies, control rods, safety rods, and instrument assemblies securely in place. 

R Reactor achieved operating status first, in December 1953, and was shut down permanently in 1964 

when the demand for the weapons reactor products began to decrease. 

P Reactor was started in February 1954 and was shutdown in August 1988 for maintenance. In 

February 1991, it was placed in cold standby and was to be used to provide spare parts for L Reactor 

and K Reactor. However, this potential use was eliminated by the subsequent permanent shutdown of 

both the L and K Reactors. 

L Reactor achieved operating status in August 1954 and was placed in cold standby in 1968. It was 

restarted in October 1985, after upgrading, and was shut down for maintenance and safety upgrades in 

August 1988. It was placed in warm standby in December 1991, to be put into operation as a backup 

to K Reactor, if necessary; but was subsequently shut down permanently. 

C-Reactor achieved operating status in March 1955 and was shut down in 1985 for maintenance. It 

was placed in cold standby in 1987 when cracks were observed in the reactor vessel. It was 

subsequently placed in permanent shutdown status. 

K Reactor achieved operating status in October 1954 and was shut down in August 1988 for 

maintenance. Initial steps to restart K Reactor began in December 1991. Successful power ascension 

testing was completed in July 1992. Following ascension testing, the reactor was taken offline to 

allow for the tie-in of a new cooling tower. The tie-in was completed and an operating permit was 

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issued in December 1992. In 1993, the cooling tower was successfully tested; however, the reactor 

was never restarted. K Reactor was then placed in cold standby, but the official status was changed to 

cold shutdown in 1996. 

In all of the reactors, some of the deuterium in the heavy water was converted to radioactive tritium, 

other radioactive materials sometime escape from the fuel elements by way of defects in the cladding. 

Accordingly, the operating crews in the reactor buildings were shielded from the high-level radiation 

within some of the reactor operating areas, and protected against airborne radioactive materials. The 

routine work spaces in the reactor buildings routinely occupied by workers, such as the offices, 

control rooms, change rooms, and shops, had an independent ventilation system. The reactor process 

areas, including the fuel element assembly rooms and spent fuel storage basins, utilized once-through 

ventilation and were maintained at lower air pressures than the non-process building areas. The 

process area ventilation air was normally discharged through a series of air filters and up relatively tall 

stacks. 

An important aspect of the SRS reactor process area ventilating systems was that they were always 

online. They needed no emergency action initiator if an abnormal level of radioactivity was detected 

in the air within the reactor rooms. The ventilation systems continued to operate regardless of the 

operating or shutdown status of the reactor. 

In the operation of the reactors, particularly during the charging and discharging of fuels and targets, 

there was inevitably some dilution of the heavy water by ordinary water present as humidity in the 

ventilating air. Efficiency of operation of the reactors depended on maintenance of the chemical and 

isotopic purity of the moderator. Accordingly, auxiliary equipment included ion exchangers for 

chemical purification and continuous vacuum distillation columns to remove light water-contaminated 

moderator for return to the 400-D area. (Bebbington, 1990; SRS Highlights, 2000). 

5.1.2 Heavy Water Rework Facility (400-D Area) 

U.S. scientists developed the Girdler Sulfide (GS) chemical exchange production process in the 1930s 

and 1940s, which was first demonstrated on a large scale at a production facility built in Dana, Indiana 

in 1945. Using the GS process, the heavy water content of ordinary water can be increased from 

about 0.015% to 15-20% by counter-current exchange of water and hydrogen sulfide gas, first through 

cold towers and then through hot towers (Bebbington, 1990). Heavy water separations operations 

using the Girdler process began at SRS in October 1952. The primary process used 144 120-foot tall 

heavy cylindrical towers ranging from 6.5 feet to 12 feet in diameter. A second stage plant, using 

vacuum distillation in smaller towers, was built at the same time as the GS facilities. A facility to 

generate steam for the heavy water distillation process, and 75-megawatts of electricity for use on the 

entire site, was also built in the 400-D Area in the early 1950s. The nominal production capacity of 

each of the two stages of the plant was designated at 240 tons of heavy water per year. This rate was 

attained in April 1953. By mid-1954, production was over 400 tons per year, and reached 500 tons 

per year in 1956 (DOE, 2008). 

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5.1.3 Defense Waste Processing Facility (DWPF) 

The Defense Waste Processing Facility processes and immobilizes the alkaline radioactive high-level 

waste (HLW) sludge generated at SRS into a durable borosilicate glass suitable for long-term storage 

in a geological repository. Following a ten-year construction period beginning in 1983, and a threeyear non-radioactive testing program, full-scale operations began in March 1996. 

Roughly 36 million gallons of liquid high-level nuclear wastes are now stored at SRS in 49 

underground carbon-steel tanks. This waste contains about 421 million curies of radioactivity, of 

which the vast majority will be vitrified at the DWPF. It is projected that the DWPF will produce 

more than 5,000 vitrified waste canisters by 2019. As of January 2008, DWPF had produced more 

than 2,000 filled waste canisters derived from over 2,000,000 gallons of HLW sludge (SRS Defense, 

2007). 

5.1.4 Saltstone Facility 

Saltstone consists of two facility segments: the Saltstone Production Facility (SPF) and the Saltstone 

Disposal Facility (SDF). Construction of SPF and the first two waste disposal vaults were completed 

between February 1986 and July 1988. The facility started radioactive operations on June 12, 1990. 

Since that time, it has been operated on an intermittent, as-needed basis to immobilize and dispose of 

low-activity liquid waste from the SRS Effluent Treatment Project (ETP) that processes waste from 

the F and H Canyons and their respective tank farms. These solutions contain low-level radioactivity 

and heavy metal ions. Immobilization is accomplished by mixing the solution with flyash, cement, 

and slag in the SPF, and pouring it into large concrete vaults to harden in the SDF. Saltstone also 

includes facilities to accommodate low-activity waste originating from the SRS Deliquifaction, 

Dissolution, and Adjustment (DDA) process, the Actinide Removal Process/Modular Caustic Side 

Extraction Unit (ARP/MCU), and eventually, the Salt Waste Processing Facility (SWPF) scheduled 

for start-up in 2013 (SRS Saltstone, 2007). 

5.1.5 Tritium Facilities 

The SRS Tritium Facilities provide tritium processing capabilities needed for post-cold-war nuclear 

weapons production and non-weapon uses. The Tritium Facilities consist of three main active process 

areas. 

The first area, built in the early 1950s, includes Building 232-H which is a 55,000 square foot facility 

that historically performed all tritium extraction and purification operations. Ancillary to Building 

232-H are Building 233-H (35,000 square feet), used for loading and unloading; and Building 234-H 

(46,000 square feet), used for shipping and receiving functions. 

The second area, called the H Area New Manufacturing Facility (HANM), began operations in 1994. 

It was significantly upgraded in 2004 to consolidate tritium processing and handling activities, 

improve safety, reduce costs, and better control environmental releases. 

The third and newest process area, the Tritium Extraction Facility (TEF), located in the H Area, 

became fully operational in early 2007. 

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The nation’s tritium production capability supporting nuclear weapons manufacturing was temporarily 

suspended in 1988 with the shutdown of the last heavy water reactor at Savannah River. In December 

1998, a decision was made that new tritium for nuclear weapons would be procured from governmentowned nuclear power reactors operated by the Tennessee Valley Authority (TVA). Subsequently, 

Tritium Producing Burnable Absorber Rods (TPBARs) were loaded into the reactor at the Watts Bar 

Nuclear Plant, located in east Tennessee. After being irradiated, the TPBARs are shipped to the 

Savannah River Site for processing at the TEF. The extracted tritium is then piped to the Tritium 

Loading Facility for further purification prior to shipment to Department of Defense (DOD) 

installations where stockpiled weapons can be replenished on an as-needed basis (SRS Spent, 2008). 

5.1.6 300 M-Area Fabrication Facilities 

Historically, the 300 M-Area was the location of facilities that produced fuel tubes of enriched 

uranium-aluminum alloy with aluminum cladding. The fuel tubes were combined into assemblies and 

loaded into the production reactors to produce Pu-238, Pu-239, Cf-252, and tritium. 

The current mission for the M Area Facilities is to facilitate the safe management of waste materials 

originating from reactor shut-down activities, to treat previously-generated mixed low-level waste in 

the Liquid Effluent Treatment Facility (LETF), and to prepare for the transition of facilities to DOE’s 

Environmental Management (EM) programs for decontamination and decommissioning (D&D). 

Starting in January 1953, SRS canned thorium metal slugs for Hanford fuel tests in Buildings 313-M 

and 320-M. In 1954, SRS began R&D and a small-scale production process to can thorium metal 

slugs for inclusion in SRS reactor fuel. By 1955, SRS was performing inspection and acceptance 

testing of thorium metal slugs that had been canned by off-site vendors. In 1964, SRS installed 

vibratory-compaction equipment in 313-M to facilitate the production of thorium oxide (thoria) slugs. 

Building 320-M was where the alloy fabrication process was conducted for target tubes and control 

rods (lithium-aluminum alloy cores with aluminum cladding). Fabrication operations have been 

discontinued and all lithium target tubes have been removed from the facility. 

Building 322-M was the Reactor Materials Quality Metallurgical and Physical Testing Laboratory 

where both fissile and non-fissile materials were examined. The building vault is now used to store 

waste materials containing fissile material. 

Buildings 330-M and 331-M were warehouses where depleted uranium cores and slugs were stored, 

but now contain only small amounts of residual waste materials. Building 340-M was a liquid 

waste-handling facility that supported the Building 322-M operations, but now contains only small 

amounts of residual contamination. 

The Liquid Effluent Treatment Facilities (LETF) consisted of three main facilities: the Dilute Effluent 

Treatment Facility [DETF] (Building 341-M); the Interim Treatment and Storage Facility [ITSF], 

(Building 341-1M); and the Vendor Treatment Facility [VTF] (Building 341-8M). The DETF was 

built in 1982 and received the production wastewater formerly discharged to the M-Area settling 

basins and outfalls. The DETF contained batch systems that processed wastewater by pH adjustment, 

precipitation, and filtration. The ITSF was built in 1985 to store and prepare production solids 

formerly disposed of in the low-level waste burial grounds. The VTF was built in 1996 to convert the 

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low-level mixed waste sludge from the ITSF into leach-resistant solid glass gems. D&D of all of the 

buildings/facilities in the LETF area has been completed (Arcano, 1994). 

5.1.7 Receiving Basin for Offsite Fuel (RBOF) and L-Basin 

The Receiving Basin for Offsite Fuel (RBOF), located in the H Area near the center of the SRS, was a 

spent-fuel storage pool for research reactor fuels. It became fully operational in 1964 and continued 

program or decommissioning operations through 2006. The mission of the RBOF was to: 

•	 receive, handle, and store irradiated nuclear fuel elements from off-site power and research 

reactors, from foreign country reactors, and from on-site reactors. 

•	 repackage nuclear fuel elements into containers and bundles for extended storage and/or shipment 

to on-site or off-site reprocessing facilities. 

•	 handle, separate, and transfer wastes generated from nuclear fuel element storage. 

From 1964 to 1988, the RBOF supported the SRS mission through the safe interim storage of 

irradiated nuclear fuel elements from U.S., off-site, and foreign reactors in support of nonproliferation 

policies. Plans were then implemented to remove the spent fuel from the RBOF and relocate this 

material to the L Area Disassembly Basin (L-Basin), a much larger facility. Following the completion 

of modifications in 1996, L-Basin received its first shipment of foreign spent fuel in February 1997. 

By October 2003, all the spent fuel previously stored in the K-Basin and the RBOF had been removed 

either to chemical separations facilities or the L-Basin, leaving the L-Basin as the only remaining SRS 

spent fuel receipt and storage facility. 

5.1.8 F Canyon 

F Canyon, one of two former chemical separations areas at SRS, is a 128,000 square foot plutonium 

and uranium separations facility previously used to process plutonium and other materials for national 

defense purposes. The facility was built in the early years of SRS operations and began chemical 

separations in November 1954. The operations area is 835 feet long, 122 feet wide, and 66 feet high, 

resembling a gorge in a deep valley between steeply vertical cliffs, hence the term “canyon.” 

To limit worker exposure to the relatively-high radiation fields in the canyon, work was remotely 

performed using overhead cranes. Thick, very dense concrete walls also separated workers from the 

radiation present in the actual interior processing areas. 

F Canyon began phasing out its traditional production operations in the late 1990s and all operations 

were concluded by March 2002. The F Canyon operations included a plutonium purification cycle to 

concentrate Pu-239 for transfer to the FB Line; processing spent fuel rods from the Taiwan Research 

Reactor (TRR) and Mark-31 targets; storing other plutonium, uranium, and Am/Cm solutions 

awaiting restart authorization; and operating nondiscretionary waste evaporation cycles to process 

canyon, analytical laboratory, reactor, and related waste streams. The F Canyon Outside Area 

Facility, including the FA Line that was used for uranium processing, was a 37,500 square foot 

complex providing support operations. 

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While awaiting a decision regarding final D&D plans, portions of the F Canyon facilities were used 

until 2008 to repackage noncompliant transuranic waste for shipment to the Waste Isolation Pilot 

Plant (WIPP) located in New Mexico (SRS F, 2008). 

5.1.9 FB Line 

FB Line, a 55,000 square foot area located in Building 221-F on top of F Canyon, was placed into 

operation in November 1954 to receive Pu-239 nitrate solution produced in F Canyon and convert it to 

a solid form (SRS at Fifty, 2002). Solutions were transferred from the canyon and concentrated in the 

FB Line. Then, in subsequent operations, the plutonium was precipitated, filtered, dried, and finally 

reduced to metal form. Process operations continued until March 2002, when all scheduled operations 

to stabilize plutonium-bearing materials from the SRS production era were completed. For about two 

years after those operations were completed, FB Line’s focus was to stabilize and package legacy 

nuclear materials for safe, long-term storage. This process involved packaging materials using a 

process in which stabilized plutonium is placed in rugged, welded stainless steel cans. After materials 

were stabilized and packaged, they were shipped to other site locations until the Mixed Oxide Fuel 

Fabrication Facility was ready. The facility is now in safe shutdown status awaiting a DOE decision 

concerning D&D (SRS FB, 2007). 

5.1.10 JB Line 

The JB Line was a two-story plutonium-finishing facility constructed on top of the F Canyon building. 

Construction was started in 1956 but not completed until 1959 (Bebbington, 1990; WSRC-MS-2000­

00061). The JB Line was built to permit high-capacity solvent extraction and plutonium finishing. 

The JB Line incorporated the trifluoride precipitation process and was built with facility safety 

improvements captured from the lessons learned in the original B Lines. Over following years, inplant process and equipment improvements were made that resulted in JB Line production that 

exceeded government requirements for quantity and quality. The highest annual production for the JB 

Line was recorded in 1983 (WSRC-MS-2000-00061). 

5.1.11 H Canyon 

H Canyon is a 403,000 square foot facility originally constructed in the early 1950s to house fissile 

isotope separations process equipment. H Canyon is 835 feet long with several elevation levels to 

accommodate the various stages of material stabilization, including control rooms to monitor overall 

equipment and operating processes, and an equipment and piping gallery for solution transport, 

storage, and disposition. Operations involving the processing of depleted uranium fuel using the 

PUREX process were initiated in July 1955. Work in the canyon, including maintenance, was 

remotely performed by overhead bridge cranes so that worker exposure to radiation could be 

minimized. The thick, dense concrete walls that separate workers from the actual processing areas 

provide added protection. 

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The facility’s operations also included chemical separations recovery of U-235 and Np-237 from 

aluminum-clad enriched uranium fuel tubes derived from SRS reactors and other domestic and foreign 

research reactors. In addition, H Canyon was equipped with capabilities to recover Np-237 and 

Pu-238 from special irradiated targets. Pu-238 was produced by irradiating recovered Np-237 in the 

SRS reactors. The Pu-238 was then recovered in H Canyon and used in power systems for the 

National Aeronautics and Space Administration’s deep space exploration programs, such as the 

Cassini spacecraft that explored the planet Saturn. 

For the historic H Canyon operations, nuclear materials (fuel tubes, oxides, etc.) were typically 

transferred from designated SRS storage areas to H Canyon, converted to solution, and then 

transferred through various process stages by which uranium, neptunium, and plutonium were 

separated. Occasional campaigns to separate U-233 and thorium were carried out in the 1960s. Waste 

material was transferred to the SRS high-level waste storage tanks for eventual vitrification in the 

Defense Waste Processing Facility. 

In 1992, the H Canyon facilities were placed in standby status; however, there remained a significant 

inventory of highly-enriched uranium fuels and solutions in various components of the SRS process 

equipment. Between December 1995 and October 1997, DOE issued a series of decisions to resume 

H Canyon chemical separation operations in order to stabilize and safely manage most of the 

remaining SRS inventory of highly-enriched uranium (HEU). DOE also concluded that H Canyon 

could be used to support stabilization of Np-237 stored in H Canyon and a number of plutonium solids 

stored in F Area vaults. 

In October 1997, H Canyon renewed operations to dissolve HEU into chemical solutions that could 

then be blended with natural uranium (NU) to form a source of low-enriched uranium (LEU) suitable 

for light water commercial nuclear power reactors. In July 2003, the first SRS LEU solution shipment 

was sent to an off-site facility operated by the Tennessee Valley Authority (TVA), which converted 

the LEU solutions into solid fuel for use in their power reactors at the Browns Ferry Nuclear Station. 

These LEU solution shipments continued through 2007. 

Areas within H Canyon have been used for other purposes. For example, the H Canyon truck well 

was used to repackage the SRS inventory of radioactive transuranic (TRU) waste into containers 

suitable for relocation to DOE’s Waste Isolation Pilot Plant (WIPP) in Nevada, which serves as 

DOE’s long-term storage facility for all TRU waste materials (Washington Savannah River Company, 

2008). 

5.1.12 Old HB Line 

The Old HB Line was a Pu-239 processing facility completed and brought online in 1953. It was 

periodically improved and upgraded throughout its operational life, and was shut down in 1984 when 

it was replaced by the New HB Line. It produced Pu-239 buttons from 1953 to 1960, and then was 

upgraded to support the NASA programs in the production of Pu-238 oxide primarily as a heat source 

for generating electricity for spacecraft going into deep space. The Old HB Line also processed 

neptunium oxide when SRS reactors were in operation. 

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Limited decontamination of the Old HB Line facility began in 1984 but was interrupted in 1986 due to 

funding restraints. At that time, the scrap recovery process equipment and the shielding for the 

neptunium oxide process equipment had been removed. The D&D efforts resumed in 1988. 

(Bebbington, 1990; DOE/EA-0948). 

5.1.13 New HB Line 

The New HB Line, is a 28,000 square foot plutonium/neptunium processing facility that was built in 

three phases during the early 1980s. The New HB Line plutonium processing facilities are located on 

top of the H-Area Canyon Building 221-H. The Frame Waste Recovery process is located within the 

221-H building. The New HB Line facility also houses a vault for the storage of Pu-238 oxide 

product and scrap material. 

There are three process lines. Phase I, or the Scrap Recovery Line, became operational in the late 

1980s and is used to dissolve and dispose of legacy plutonium scrap. It is also used to dissolve legacy 

enriched uranium for blending into low-enriched uranium to be shipped to the Tennessee Valley 

Authority for fabrication into commercial power reactor fuel. The Phase I process converts solid 

nuclear materials into nitrate solutions and transfers those solutions to H Canyon for disposition. 

Phase II, or the Np-237/Pu-239 Oxide Line, can produce oxide (powder) material from Np-237 or 

Pu-239 nitrate solutions. Phase II started operations for the first time in November 2001. The 

plutonium material was shipped to FB Line for packaging in 3013 containers for long-term storage, 

and then to K Area for storage. The neptunium material is shipped to the Idaho National Laboratory 

for further processing and conversion to reactor targets for future Pu-238 production. Phase III, or the 

Plutonium-238 Oxide Production Line, was converted into a processing facility to open storage 

containers when necessary, and oxidize metals to allow for dissolution in the Phase I process area. 

Phase III supports missions to disposition legacy plutonium and uranium metals and oxides 

(DOE/EA-0948). 

5.1.14 Naval Fuels Manufacturing Facility 

The Naval Fuels Manufacturing Facility in Building 247-F was a two-story, 110,000 square foot 

enriched uranium fuel manufacturing facility built in the early 1980s. It operated from 1985 through 

1989 to provide additional naval nuclear fuel manufacturing capacity for the cold war effort. The 

manufacturing processes employed a wide variety of acids, bases, and other hazardous materials. As 

the cold war wound down, the need for naval fuel declined. Consequently, the facility was shut down 

and underwent initial deactivation. All process systems were flushed with water and drained using the 

existing process drain valves. However, since these drains were not always installed at the lowest 

point in piping and equipment systems, a significant volume of liquid remained after initial 

deactivation was completed in 1990. At that time, a non-destructive assay of the process area 

indicated that as much as 34 kg of uranium might remain in equipment and piping. The 247-F 

Closure Project is currently underway that will result in the final D&D of the Naval Fuels 

Manufacturing Facility. 

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5.1.15 Building 235-F Plutonium Fuel Fabrication Facility (PUFF) 

The 235-F Plutonium Fuel Fabrication Facility (PUFF), or Building 55, was a 55,000 square-foot 

operations facility containing Pu-238 hot cells. Operations to produce encapsulated Pu-238 oxide fuel 

forms and, on occasion, neptunium billets began in 1978. In December 1983, DOE completed Pu-238 

fuel clad production for NASA's Galileo and Ulysses space missions at PUFF, after which the 

equipment was placed on operational standby status. Currently, the PUFF equipment is being used to 

provide safe storage of residual nuclear material resulting from past nuclear weapons production and 

other non-weapon uses that are no longer required. 

5.1.16 Savannah River Technology Center 

The Savannah River Laboratory research facilities include the main laboratory building, 773-A; an 

experimental physics laboratory, 777-M; the CMX and TNX process pilot facilities located near the 

400-D Area; the Health Physics Laboratory, 735-A; and the equipment engineering laboratory and 

shops, 723-A. CMX was closed down in 1983 when its function (long-term flow testing of new fuel 

and target elements) was moved to 773-A. The 777-M operations were also discontinued in the 1980s 

when modern computers eliminated the need for many of the experimental measurements that were 

previously needed to support reactor charge design. 

5.2 Radiological Exposure Sources from Savannah River Site Operations 

ATTRIBUTION: Section 5.2 and its related subsections were completed by Ray Clark, Oak Ridge 

Associated Universities; Bryce Rich, Mel Chew and Associates, Inc.; Sam Chu, Mel Chew and 

Associates, Inc.; and Eugene Potter, Mel Chew and Associates, Inc. These conclusions were peerreviewed by the individuals listed on the cover page. The rationales for all conclusions in this 

document are explained in the associated text 

SRS operations involved several processes of the nuclear weapons development cycle (DOE, 1997; 

DNA-1251-1-EX) and played a significant role within the US nuclear weapons program. These 

processes include nuclear fuel fabrication; nuclear reactor operations; radiochemical separations; 

radionuclide production - including nuclear weapons materials; recycling uranium; refining, finishing 

and storing plutonium; and handling the associated radioactive waste. The following subsections 

provide an overview of the internal and external exposure sources associated with these operations for 

the class under evaluation. 

5.2.1 Internal Radiological Exposure Sources from SRS Operations 

At SRS, tritium produced most of the personnel exposure from internal deposition. Major sources of 

non-tritium intakes were potential uptakes of uranium and plutonium; secondary sources were mixed 

fission products and activation products. 

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Table 5-1 provides a summary of radionuclides that are recognized as potentially contributing to 

internal dose to SRS workers. These radionuclides of concern (ROCs) are listed in Chapter 5 of the 

SRS Internal Dosimetry Technical Basis Manual (WSRC-IM-90-139). These ROCs were identified 

by SRS as the most significant internal intake sources, or effective tracers for evaluating the 

significant dose-delivering radionuclides (ORAUT-TKBS-003, p. 64). The following subsections 

address these ROCs in turn. 

Table 5-1: Potential Contributors to SRS Internal Dose 

Radionuclide 

H-3 

Pu-238, Pu-239, Pu-240, Pu-241, Pu-242 

U-233, U-234, U-235, U-236, U-238, and mixtures 

Np-237 

Am-241, Am-243 

Cm-244 

Natural thorium 

Mn-54 

Co-60 

Zn-65 

Sr-90 / Y-90 

Ru-106 

Sb-125 

Cs-134, Cs-137 

Ce-144 

Eu-152, Eu-154 

Tm-170 

Cf-252 (including Cm-248 daughter) 

5.2.2 External Radiological Exposure Sources from SRS Operations 

SRS was built to produce the basic materials used in the fabrication of nuclear weapons, primarily 

tritium and Pu-239. Five reactors were built to produce nuclear materials (including nuclear weapons 

materials) through neutron irradiation of target materials in the production reactors. Support facilities 

included two chemical separations plants, a heavy water extraction plant, a nuclear fuel and target 

fabrication facility, and waste management facilities. 

The fuel elements used in the reactors generated direct beta, photon, and neutron radiation fields, as 

well as the potential for surface and airborne contamination by uranium, TRU isotopes, and a wide 

variety of fission and activation products. Separations processes, research and development activities, 

and the collection and disposal of radioactive wastes also resulted in potential direct exposures to 

direct, beta, photon, and neutron radiation fields. 

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Neutron exposure sources consisted of: (1) the operating reactors; (2) the separations processes and 

product lines, which handled quantities of TRU isotopes and the associated spontaneous fission and 

α/n neutrons produced; (3) the analytical and production control laboratories to a lesser degree; (4) 

R&D facilities, which also handled TRU isotopes in significant quantities and associated neutron 

emissions; (5) vaults and TRU storage facilities; and (6) calibration facilities and others whose 

operation relied upon the use of neutron sources. 

5.2.3 Incidents 

NIOSH has found no evidence or documentation of any incidents that would have resulted in very 

high exposures (similar to those received in a nuclear criticality accident). 

Starting in March 1954, some incidents involving work with radioactive materials were documented, 

and worker exposures were assessed when incidents involved radioactive sources (SRS Dosimetry, 

1959; SRS Dosimetry, 1962; SRS Dosimetry, 1974a; SRS Dosimetry, 1974b; SRS Dosimetry, 1989; 

DPSP-55-454-2). Reports of these incidents were retained in SRS Dosimetry Special Hazards 

Investigations files. Copies of many of these reports have been obtained through 1989 and are in the 

SRDB. SRS Special Hazards bulletins required Construction supervision to be part of incidents 

involving radioactive materials and construction workers (DPSOP-40, pdf pp. 71, 183, 248). 

NIOSH completed a search of the SRS Site Incident database that lists 332 exposure incidents from 

1954 through 1996. The search found that 481 worker incident records with “urine results.” Further 

in-depth investigation revealed 31 workers received confirmed intakes of radioactivity. Also listed in 

the incident reports were 12 neutron exposures and 12 external exposures (Singh, 2007). NIOSH 

found incident monitoring data in NOCTS for former SRS workers who were identified in incident 

records. 

Construction workers were involved in some of the recorded incidents beginning in 1954. In the first 

100 incidents from the SRS Dosimetry Special Hazards Investigations files (March 1954 – March 

1959), construction workers were involved in twenty-six incidents. In each incident, workers were 

monitored for contamination. Depending upon the type of incident, they were also monitored for 

external dose and for internal uptake of radioactivity. 

6.0 Pedigree of Savannah River Site Data 

This section answers key questions prior to 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. 

Historical information gathered for SRS indicates that the management of individual radiation 

exposure records began at hiring with the accumulation of data pertinent to the individual and any 

previous occupational radiation exposure. Appropriate radiation monitoring badges or devices were 

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issued and accountability was established (DPSPU-75-30-7). Site dosimetry records have been 

maintained by the same organization since the site’s inception. Employee identification numbers were 

used as the sole tracking method for the early years. In the early 1960s, the practice of re-using the 

identification numbers of terminated employees was begun, which complicated the unique tracking of 

an employee’s historical exposure. Social security numbers were included in the dose records in 

1973. From 1951 to 1957, personnel exposure was recorded on various data forms and stored in 

individual file folders. From 1958 through 1972, employee exposure information was sent to a central 

computing facility for entry into a computer file. Quarterly computer-generated compilations of 

exposures for the cycle, quarter, year, and site were maintained in “logbooks.” 

In 1973, a more sophisticated computer program, HP Master File, replaced the previous system. Each 

year’s records were stored on a single magnetic tape. The HP Master File also produced bioassay 

sampling schedules, whole body/chest count schedules, and tritium sample results for monthly dose 

equivalent calculation as well as maintaining accountability for film badges. 

In 1979, a computerized system to produce annual dose reports to employees (“report cards”) was 

initiated. The Health Protection Annual Radiation Exposure History (HPAREH) system was 

developed by building and verifying a history file of annual radiation exposure data from handwritten 

data, logbooks, and magnetic tape. The first report cards were issued in 1980 (for 1979), and the 

HPAREH database was updated from the HP Master file for each subsequent year. 

In 1989, the HP Master File system was replaced by the Health Protection Radiation Exposure 

Database (HPRED) system. Previously, data were either hand-entered on computer input sheets or 

processed on magnetic tapes and then sent to the site’s computer group. With HPRED, data was input 

either using terminals or directly from the TLD badge reader system. In 2004, HPRED was replaced 

with a commercial product (ProRad). From 1951 to 1983, visitor exposure records were posted 

manually on individual 3-inch by 5-inch cards and filed by year in a card file. The cards were 

separated into plant or construction visitors and placed in alphabetical order (WSRC-RP-95-234). 

6.1 Internal Monitoring Data Pedigree Review 

ATTRIBUTION: Section 6.1 and its related subsections were completed by Sam Chu, Mel Chew and 

Associates, Inc.; and Eugene Potter, Mel Chew and Associates, Inc. These conclusions were peerreviewed by the individuals listed on the cover page. The rationales for all conclusions in this 

document are explained in the associated text. 

NIOSH has found that SRS policies for the collection and maintenance of employee monitoring data 

are sufficient for dose reconstruction in terms of the pedigree parameters described above. Upon 

request, the SRS provides database printouts and copies of original hardcopy bioassay and whole body 

count/chest count records. For most of the site’s history, the data are provided as originally recorded, 

but since 1990, excreta analysis records have been maintained in an electronic database and computergenerated reports are provided. 

Since no electronic data are available before 1990, data from NOCTS have been entered into a 

database for the purpose of creating a co-worker model for workers who were unmonitored or for 

whom records are unavailable. There are approximately 382,000 in vitro bioassay records available 

for analysis. After entry, these data were subject to 100% verification by a second person. NIOSH 

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has access to logbooks containing bioassay starting in 1954. A comparison of the computerized 

printout sheets and logbook entries for a number of claimants was completed to establish credibility 

and consistency of the internal dosimetry data. No evidence of systematic errors or significant 

differences between the printouts and hardcopy data was observed. 

NIOSH reviewed entries in four bioassay logbooks covering a period of six years as a representative 

sample of the data. Of the 200 logbook entries reviewed, 62 were claimants in the NOCTS database. 

Three claims contained no data corresponding with the logbook entries (< 5%). Fifty-seven claims 

had corresponding data (92%). Forty-two percent of the claimants were in construction-related 

positions. 

6.2 External Monitoring Data Pedigree Review 

ATTRIBUTION: Section 6.2 and its related subsections were completed by Bryce Rich, Mel Chew and 

Associates, Inc.; Eugene Potter, Mel Chew and Associates, Inc; and Darin Hekkala, Dade-Moeller, 

Inc. These 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. 

Records of radiation dose to individual workers wearing personnel dosimeters are available for SRS 

operations beginning in 1951. Dose from these dosimeters was recorded at the time of measurement 

and routinely reviewed by SRS operations and radiation safety staff for compliance with radiation 

control limits. Based on the SRS documentation, there does not appear to be any significant 

administrative practice that would affect the integrity of the recorded doses for SRS workers with the 

exception that some workers exposed to natural uranium might not have been monitored prior to July 

1955. 

For monitored workers, NIOSH has obtained the complete set of external dosimetry records on the 

quarterly reports since 1958, including the visitor cards. Although this is the complete dataset, the 

primary source of the co-worker dose model in ORAUT-OTIB-0032 was the Health Protection 

Annual Radiation Exposure History Database (HPAREH) developed in 1979. This database consists 

of annual doses to workers (shallow, deep, neutron, and tritium) and was used from 1980 to 1989 to 

produce the annual reports. Some workers who terminated employment prior to 1979 and visitors are 

not included in HPAREH. However, external exposure records for these employees are available on 

paper, either in the employee’s individual dosimetry file or in visitor records. Table 6-1 presents the 

number of annual dose records contained in the HPAREH database for SRS workers along with the 

total number of monitored workers listed in WSRC-RP-95-234. 

In April 1989, the Health Protection Radiation Exposure Database (HPRED) was implemented. This 

system interfaced directly with the TLD badge reader system. In 1990, visitor records were included 

in this database. 

In 2004, HPRED was replaced by a commercial access control and radiological records management 

system, the ProRad database. Advantages of the system over the previous one included: real-time 

availability of individual actual and estimated dose totals, automated personnel qualification checks, 

improved efficiency and real-time availability of radiological information (SRS Intake, 2004). 

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Table 6-1: Summary of Available Monitoring Data for SRS Workers 

Year 

No. of 

Monitored 

Workers per 

WSRC-RP95-234 

Total Number of Annual External Dose Records in 

HPAREH 

No. of 

Monitored 

Workers 

Shallow Dose 

Records 

Deep 

Dose 

Records 

Neutron 

Dose 

Records 

1950 2 0 0 0 

1951 50 29 0 0 0 

1952 400 270 177 177 0 

1953 1700 992 594 594 1 

1954 4800 2736 1391 1391 1 

1955 5500 3177 2012 2012 1 

1956 5600 3235 2740 2740 10 

1957 5900 3375 2971 2971 19 

1958 6200 3500 3132 3132 17 

1959 5756 3189 2804 2804 7 

1960 5670 3425 3253 3253 17 

1961 5819 3537 3358 3358 10 

1962 6332 3630 3474 3474 17 

1963 6219 3815 3501 3501 33 

1964 5498 3814 3372 3372 120 

1965 4977 3682 3468 3468 178 

1966 5032 3720 3329 3329 184 

1967 5042 3819 3603 3603 223 

1968 4875 3822 3670 3670 365 

1969 4705 3797 3671 3671 517 

1970 5224 3810 3412 3412 484 

1971 4856 3956 3302 3302 597 

1972 4737 4175 3384 3384 531 

1973 5639 4369 3534 3534 462 

1974 5653 4472 3809 3809 480 

1975 6129 5056 4329 4329 424 

1976 6336 5456 4784 4784 486 

1977 6318 6164 5057 5057 517 

1978 7647 7067 5760 5760 403 

1979 7615 7763 6286 6286 502 

1980 8066 8213 6571 6571 675 

1981 8951 9131 6764 6764 840 

1982 9192 9371 6515 6515 608 

1983 10195 10396 7393 7393 670 

1984 11740 12006 7074 7074 781 

1985 13950 14253 8323 8323 935 

1986 14236 14544 10949 10949 802 

1987 14096 14565 11700 11700 889 

1988 15458 15920 12438 12438 787 

1989 17468 17925 13674 13674 733 

1990 20535 20934 11350 11350 697 

1991 20428 20832 19908 19908 2135 

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Table 6-1: Summary of Available Monitoring Data for SRS Workers 

Year 

No. of 

Monitored 

Workers per 

WSRC-RP95-234 

Total Number of Annual External Dose Records in 

HPAREH 

No. of 

Monitored 

Workers 

Shallow Dose 

Records 

Deep 

Dose 

Records 

Neutron 

Dose 

Records 

1992 19320 19682 18734 18734 2083 

1993 16747 17078 16077 16077 2926 

1994 13983 14120 12587 12587 3057 

1995 (see Note) 13923 12450 12450 3385 

1996 N/A 12716 11400 11400 3414 

1997 N/A 12293 11069 11069 3382 

1998 N/A 11620 10484 10484 3261 

1999 N/A 11150 10095 10095 3306 

Note: Data values in Column 2 end at 1994 because WSRC-RP-95-234 was published in 1995. 

Figure 6-1: Comparison of the Number of Monitored Workers in HPAREH vs. WSRCRP-95-234 and Hardcopy 4th Quarter Reports Obtained from SRS in August 2008 

As mentioned above, not all historical monitoring data was stored in the HPAREH database. The data 

for some workers who terminated employment prior to 1979 and visitors are not included in HPAREH 

since these data was not needed to produce the annual reports. Thus for pre-1979 data, a comparison 

of the number of workers in the hardcopy quarterly reports with HPAREH and the data in WSRC-RP95-234 was conducted. As illustrated in Figure 6-1, for 1960 approximately 53% of the monitored 

workers are in HPAREH. The percentage of workers included in HPAREH generally increased with 

time till 1979 when all workers were included. 

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The petitioners directly questioned the ability to reconstruct doses in light of this missing data in 

HPAREH. For this evaluation, a review of the annual dose distributions was conducted in five-year 

increments between 1960 and 1975 (Figure 6-2). Based on this evaluation, the annual dose 

distributions using all of the monitoring data in the 4th Quarter reports were less than the dose 

distributions in HPAREH. As a result, even though HPAREH does not contain all of the monitoring 

data, the annual dose values used in ORAU-OTIB-0032 may be used to bound dose. 

The lower border of each bar is the 25th percentile of the distribution, the solid line midway in each 

bar represents the median, the dashed line in each bar is the mean of the distribution, the upper border 

of each bar is the 75th percentile, the upper whisker is the 90th percentile, and the circle above each 

bar is the 95th percentile of the distribution. 

Figure 6-2: Comparison of Annual Dose Distributions for Construction Trades
 

Workers in HPAREH and the 4th Quarter Reports
 

A subsequent comparison was conducted evaluating only the construction worker data and the 4th 

Quarter Reports. Table 6-2 presents the total number of construction trades workers in HPAREH and 

the 4th Quarter Reports. In general, less than half of the Construction Trades workers were listed in 

HPAREH compared to the 4th Quarter Reports. As with the all worker comparisons, HPAREH annual 

dose distribution was generally higher than the 4th Quarter Reports, as shown in Figure 6-3. The 

exception was 1970 in which the 4th Quarter Reports indicated a slightly higher 75th percentile. 

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Table 6-2: Comparison of Construction Trades Workers to HPAREH 

Year Number of Monitored Construction Trades Workers 

HPAREH 4th Quarter Reports 

1960 202 747 

1965 236 531 

1970 211 312 

1975 568 1011 

Figure 6-3: Comparison of Annual Dose Distributions for Construction Trades
 

Workers in HPAREH and the 4th Quarter Reports
 

The nature of construction work is periodic; construction workers would enter and leave before the 

annual reports were produced. To further evaluate the effect of this periodic work on the annual dose 

distributions, a complete roster of all construction workers was obtained for 1960. While HPAREH 

only listed 202 Construction Trades Workers, a total of 747 were listed on the 4th Quarter Annual 

Report. When the annual roll was reviewed which combined the 1st, 2nd, 3rd, and 4th quarters, a total 

of 1399 Individual Construction Trades Workers were identified, nearly double the 4th quarter results 

and 7 times the number in HPAREH. The 1960 annual doses for these workers were coded and 

compared (see Figure 6-4). As with the other distributions, as more workers were included, the 

annual dose distributions decreased. The likely cause of this bias is that more of the experienced 

crafts were possibly employed for a longer period of time and used for more of the higher exposure 

jobs. This effectively resulted in a slight claimant favorable bias in the HPAREH database. 

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Figure 6-4: Comparison of 1960 Annual Dose Distributions Using the Annual Roll, 

4th Quarter Reports, and HPAREH 

Based on the review of the annual dose distributions, and specific evaluation of the construction trades 

workers, NIOSH concludes that the HPAREH database used in ORAU-OTIB-0032 is sufficiently 

robust. For the purpose of this evaluation, this pedigree assessment supports the use of these external 

monitoring data for determining the feasibility of bounding external dose for the class under 

evaluation when ORAUT-OTIB-0052 guidance is applied to the co-worker model. 

Neutron Dose 

Since 1972, the SRS has used Thermoluminescent Neutron Dosimeters (TLNDs) to measure 

personnel neutron doses at SRS. Since TLNDs are recognized as state of the art dosimetry for 

neutrons, neutron doses in HPAREH post 1972 are considered sufficiently robust for the development 

of a co-worker model. Thus, further evaluation of the pedigree of these measurements is not 

warranted. 

However, prior to 1972 the neutron doses are significantly limited due to the use of NTA film. In 

order to use this NTA data, correction factors must be applied to correct for under-response due to 

energy limitations. Based on a review of SRS reports recently captured as part of this SEC evaluation, 

SRS conducted direct measurements comparing the new TLND, area survey measurements and NTA 

film (DP-MS-69-004, 1969). Table 6-3 provides the direct work place comparison at the SRS 

Plutonium Finishing Area. The average under-response of the NTA film was approximately a factor 

of four. 

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Table 6-3: Comparison of Bonner Sphere, TLND, and NTA Film Measurements at 

the SRS Plutonium Finishing Area 

Location 12-in Bonner Sphere 

(mrem) 

TLND 

(mrem) 

NTA* 

(mrem) Correction Factor 

Plutonium 

Finishing Area 80 

89 

31 2.6 

26 3.1 

14 5.7 

90 

19 4.2 

19 4.2 

14 5.7 

Average 80 89.5 20.5 3.9 

* The NTA Film was calibrated using a high energy PuBe source. 

As noted in the SRS TBD, personnel monitoring for neutrons was only required when workers entered 

a radiation area where the neutron dose rate exceeded 1 mrem/hr. Thus, some unmonitored exposure 

to neutrons was likely. This is the primary reason the TBD uses a neutron-to-photon ratio (N:P) to 

assess the unmonitored neutron dose. 

From an external dose pedigree standpoint, validation of the neutron-to-photon ratio listed in the TBD 

will require significant work (as was required for the Hanford SEC evaluation). This work will be 

conducted after this report is submitted. Currently, the complete set of personnel neutron dose 

measurements has been reviewed on site by the SEC evaluation team and has subsequently been 

requested from SRS. As of this writing, these data are currently undergoing classification review and 

have not been formally evaluated. 

During the on-site review of the NTA data, NIOSH noted that prior to 1961 there appears to be much 

less NTA data available for verification of the N:P ratio. Due to this lack of data, NIOSH also 

reviewed Radiation Survey Reports to determine if workplace neutron measurements were conducted 

in parallel with photon survey measurements. NIOSH has verified that this in fact occurred. The 

paired measurements have been recorded on individual Radiation Survey Log Sheets (RSLS) in the 

SRS records holdings. The combination of the corrected NTA data along with the survey data will be 

used to verify the N:P ratio in the SRS TBD. This methodology is discussed in more detail in Section 

7.2. 

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7.0 Feasibility of Dose Reconstruction for the Class Evaluated by 

NIOSH 

The feasibility determination for the class of employees under evaluation in this 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. 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 assure that NIOSH can bound either the maximum doses that 

members of the class might have incurred, or more precise quantities that reflect the variability of 

exposures experienced by groups or individual members of the class as summarized in Section 7.6. 

This approach is discussed in OCAS’s SEC Petition Evaluation Internal Procedures which are 

available at http://www.cdc.gov/niosh/ocas. The next three major subsections of this Evaluation 

Report examine: 

• The feasibility of reconstructing internal radiation doses. (Section 7.1) 

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

• The bases for petition SEC-00103 as submitted by the petitioner. (Section 7.3) 

Members of the proposed SEC Class were DuPont employees plus members of private contractor 

organizations and were provided identical Health and Safety coverage as were other operations and 

maintenance workers. Health Physics support of Construction activities began in 1953. Most workers 

were provided dosimetry, protective clothing and equipment, and needed bioassay coverage when 

they entered a radiologically-controlled work area. Health Physics services provided to construction 

workers are discussed in DPSP-55-454-3 and DPSOP-40. By procedure, all radiation and 

contamination control procedures applied to “All departments and Construction” (DPSOP-40). 

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7.1 Evaluation of Bounding Internal Radiation Doses at Savannah River Site 

ATTRIBUTION: Section 7.1 and its related subsections were completed by Mike Mahathy, Oak Ridge 

Associated Universities; Eugene Potter, Mel Chew and Associates, Inc; Sam Chu, Mel Chew and 

Associates, Inc.; Robert Morris, Mel Chew and Associates, Inc.; and Will McCabe, MJW 

Corporation. These 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. 

To evaluate the petitioner’s concerns regarding internal exposures to special radionuclides produced at 

SRS, NIOSH investigated the source term, available personnel monitoring records, and air sample 

data, as needed. Based on this information, NIOSH made a determination as the feasibility of 

reconstructing internal doses for each radionuclide. 

The principal source of internal radiation intakes for members of the class under evaluation was 

tritium. Major sources of non-tritium intakes were potential uptakes of uranium and plutonium; 

secondary sources were mixed fission products, activation products, and thorium (ORAUT-TKBS0003). 

The majority of SRS workers was monitored for internal intake of radionuclides and has internal 

monitoring data. This point is reflected in the NOCTS claimant files for which 1467 of the 1798 

construction worker claim files contained internal monitoring data. However, the possibility exists 

that some workers who should have been monitored were not, or that the data for some workers who 

were monitored has been misplaced. As a result, NIOSH is developing a co-worker model based on 

the claimant data in NOCTS. There are approximately 382,000 in vitro bioassay records available for 

analysis in NOCTS. A separate study has been completed that established the principle that, under 

certain conditions, the data in NOCTS are representative of a site’s population generally (ORAUTOTIB-0075). All of the SRS NOCTS data has been entered into a database, and the co-worker model 

is in preparation. In addition, ORAUT-OTIB-0052 indicated that construction trades workers had 

more plutonium bioassay measurements below the reporting limit compared non-construction 

workers. ORAUT-OTIB-0052 also found that for the positive bioassay, the non-construction workers 

results were generally higher than construction trades workers. 

A comparison of isotope production at SRS versus internal monitoring for intakes of those isotopes is 

presented in Figure 7-1 and Table 7-1. Note the bioassay start dates in Figure 7-1 correspond to the 

start dates of bioassay in NOCTS and not necessarily the site. 

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[1] Source: Savannah River Site at Fifty (SRS at Fifty, 2002, p. 453), table entitled Isotope Production at SRS over 

Time. 

[2] This timeline (black bar) includes the first production of an isotopic heat source at SRS in 1958. This program 

continued until 1986 (pp. 357, 429). 

[3] This timeline includes high-activity Co-60 generation for isotopic heat sources that began in May 1955 (pp. 356, 

432). 

[4] This timeline includes thorium irradiation as part of Eisenhower's Atoms for Peace program with the first 

production of U-233 in 1955-56 (p. 350) 

[5] This timeline includes the Curium I and II, High Flux, and Transplutonium I and II programs that produced 

Pu-242, Am-243, and Cm-244, including Cf-252, Es-255, Fm-257, Fm-258, and Co-60 from the High Flux 

Program (pp. 357, 432, 433). 

[6] This timeline includes the Californium I program that began in August 1969 and continued until November 1970 

and produced milligrams of Cf-252 (p. 433). 

Figure 7-1: SRS Isotope Production vs. Internal Monitoring for Intakes 

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Table 7-1: Bioassay Start Dates for Specific Isotopes and Suitable Bioassay Technique 

Production 

Isotopes 

Bioassay 

start 

Logbook 

Ref. Box 

SRS ID TBD 

Rev. 8, 2001 

SRS ID 

TBD Page 

Bioassay Technique 

Thulium-170 1954 (FP) M270-4639-3 beta emitter, 84 keV 

gamma @ 3%. Gross Beta 

Europium-152 1960 (WBC) in vivo 5-12 whole body count or chest 

count 

Californium-252 1965 

(TKBS)

 urinalysis or in 

vivo 

5-3 gross alpha or chest count 

Polonium-210 2/7/1967 QH 211-1217­

94-006 

gross alpha 

Special programs, 

other isotopes 

2/16/1961 QH 211-1217­

94-006 

Isotope specific 

Plutonium-242 Gross alpha gross alpha 

Curium-244 11/19/1963 QH 211-1217­

94-005 

urinalysis or in 

vivo 

5-3 gross alpha or chest count 

Americium-243 1956 QH 211-1217­

94-006

 5-4 gross alpha 

Plutonium-238 4/5/1954 QH 211 0886­

93-015 

urinalysis or in 

vivo 

5-5 gross alpha, alpha 

spectroscopy, chest count 

Plutonium-240 4/5/1954 QH 211 0886­

93-015 

urinalysis or in 

vivo 

5-6 gross alpha, alpha 

spectroscopy 

Cobalt-60 (IA) 5/1956 QH 211-1217­

94-006 

in vivo 5-15 whole body count 

Uranium-233 Gross alpha gross alpha 

Tritium 7/29/1953 M270-4639-3 urinalysis 5-15 gross Beta 

Plutonium-239 4/5/1954 QH 211 0886­

93-015 

urinalysis or in 

vivo 

5-6 gross alpha, alpha 

spectroscopy 

Neptunium-237 12/6/1953 QH 211-1217­

94-006 

urinalysis or in 

vivo 

gross alpha, whole body 

count or chest count 

Thorium-232 urinalysis or in 

vivo 

NIOSH has obtained the hardcopy SRS bioassay logbooks dating back to 1954. Since 1990, excreta 

analysis records have been maintained in an electronic database. The database identifies baseline 

(new hire), routine, special, and termination measurements. Prior to the electronic database, excreta 

results were kept on employee bioassay cards. Chemical separation and alpha counting were 

generally used prior to 1994. NIOSH has obtained alpha analysis procedures for determination of 

plutonium, uranium, and neptunium (DPSOP-47); polonium alpha counting; and thorium alpha 

counting. The urinalysis for trivalent actinides (americium, curium, and californium), dating back at 

least to the mid-1960s, consisted of directly plating the sample and gross alpha counting. The 

plutonium urinalysis program began in 1954. Approximately 200 samples were analyzed for thorium 

in 1956. Chemical separation and alpha spectroscopy have been used since 1994 to identify specific 

isotopes (e.g., U-234, U-235, and U-238). 

Tritium urinalyses were available from start-up, but increased when SRS implemented liquid 

scintillation counting in 1958 (Health Physics Aspects, 1958). Tritium results in the 1990s were listed 

on the same summary form as external dose monitoring results. 

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Uranium urinalyses were also available from start-up. SRS technical documentation indicates that for 

earlier monitoring periods, the designations “enriched” and “depleted” reported for uranium analysis 

referred to analysis performed by alpha counting or chemical measurement, respectively, and were not 

necessarily indicative of the degree of uranium enrichment. “EU” was the code used on employee 

bioassay cards for the gross alpha count method, and “DU” was used to designate the 

fluorophotometric method. Activity fractions for depleted, recycled, and highly-enriched uranium are 

listed in the TBS. Neptunium analyses were implemented by 1959. (ORAUT-TKBS-0003, pp. 65­

71) Thorium urinalyses were performed starting in 1956. 

Intakes of gamma-emitting fission products were monitored by urinalysis, beginning in 1954 when the 

reactors went on line. The start date for urinalysis for strontium radioisotopes was December 1954, 

the month following the first separations work in F Canyon (Bioassay Logbook, 1954-57). However, 

Boni discusses a bioassay procedure in use at the time for rapid estimation of the total amount activity 

of beta-gamma-emitting isotopes in a urine specimen with recoveries of Sr-89, Sr-90/Y-90, 

Zr-95/Ni-95, Ce-144/Pr-144, Fe-59, Cr-51, and Zn-65 greater than 90% and Co-60 of 85% (DPSPU58-030-011). 

Whole-body counting has been performed since 1960 when the SRS Whole Body Counting Facility 

was completed. The first chest counting was with sodium iodide (NaI) detectors in 1965. Chest 

counting for low-energy photons was improved in the early 1970s using phoswich detectors. In vivo 

count records are not kept in a central electronic database; each count is maintained as a single 

hardcopy record. A whole-body count record and a chest-count record on the same day are listed on a 

single hardcopy record (ORAUT-TKBS-0003, p. 71). 

Air sample records were considered workplace monitoring records as opposed to intake monitoring 

records, and hence, were not maintained in the same records collection as bioassay records. Air 

sample records were associated with each facility separately and were stored as facility records 

(ORAUT-TKBS-0003, p. 75). The air sample results were recorded on Air Sample Log Sheets 

(ASLS). NIOSH has located several hundred boxes of the records through a search of the SRS 

EDWS records system. 

Historically, SRS had an extensive radiation safety monitoring program to measure radiation exposure 

in the workplace, including contamination surveys (ORAUT-TKBS-0003, p. 89). Radiation surveys 

are recorded on the Radiation Survey Log Sheets (RSLS). The RSLS are filed by facility and time. 

These records are sequentially numbered so it can be determined with certainty if some are missing. 

However, like air monitoring data, these records are not likely to be found in the individual 

employee’s records. 

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7.1.1 Radionuclides of Internal Concern 

Each radionuclide of internal concern is discussed in the following subsections. A discussion of 

personnel monitoring and a determination of feasibility of dose reconstruction is presented for each. 

7.1.1.1 Tritium 

Tritium has been one of the main production materials throughout the life of the Savannah River 

reactors. Historically, tritium resulted in personnel intakes because of the large quantities present at 

SRS. Tritium does not occur in abundance in nature but can be produced by bombarding Li-6 targets 

with neutrons to create the unstable Li-7 isotopes which decay to become tritium through alpha 

emission within the reactors. Tritium-containing irradiated targets were then transported to the 200-F 

Area initially and the 200-H Area Tritium Facility for purification and packaging into usable form. 

The production of tritium began in 1953 and continues to the present time. Between October 1955 

and 1964, SRS produced the majority of the tritium in the US. SRS has produced many kilograms of 

tritium during its production period (SRS at Fifty, 2002). 

The first production reactor started in December 1953. Tritium concentrations in the moderator 

steadily increased with reactor operation. An April 1954 HP Area Survey Report stated for the 100R 

Area “The P-10 content of the process water has reached a level such that breaking of process water 

lines requires a Scott Air Pak.” (Health Physics, 1954) A July 24, 1954 report identifies that a 

working level = 1E-5 uc/cc for tritium and that the “only positive bioassay results obtained for tritium 

to date are for two men who cleaned up a spill in E Plant... Bioassay results were above the sensitivity 

limit but well below the permissible body burden.” (Summary of Tritium, 1954) The reported average 

tritium concentration in the five reactors was 1,000 μCi/mL in January 1956, 2,000 μCi/mL in January 

1957, and 3,000 μCi/mL in December 1957. The number of tritium air samples >1E-5 μCi/mL 

(requiring job restrictions such as limiting stay time and or personal protective equipment) was 14 in 

January 1956, 173 in January 1957, and 217 in December 1957. There were four significant tritium 

bioassay results (>1 mCi assimilation) in 1956 and 49 in 1957 (Health Physics Aspects, 1958). There 

were no significant bioassay results prior to 1956 (DPSPU-58-030-014). 

As of August 26, 1957 all personnel were subject to requests for bioassay samples by Health Physics, 

as dictated by results of the survey program. Personnel subjected to air activity above prescribed 

levels were required to submit samples as requested by Health Physics. In addition to the above, all 

construction personnel who worked during shutdowns were to submit bioassay samples, as follows: 

a.	 At the end of the first week, welders and grinders, pipefitters and/or other personnel who have 

worked in close proximity to open water lines, for tritium urinalysis. 

b.	 Welders and grinders at completion of their work in the area, but prior to leaving the area, for 

induced activities (750 mL). NOTE: Sample should be started soon enough for completion prior 

to leaving the area (e.g., approximately ten days before). 

c.	 Personnel at completion of shutdown or upon completion of work in the area, but prior to leaving 

the area, for tritium analysis (Health Physics, 1957). 

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Tracking and follow-up procedures were in place to obtain samples from workers who did not submit 

a sample, as evidenced by a memo to M.W. Hartnett requesting 750 mL bioassay samples from over 

100 Pipe and Boilermaker crafts personnel who failed to provide samples following the 100-K Area 

shutdown (Bio-Assay Samples, 1957). 

Tritium was recovered in the tritium processing facilities (200 Areas). 232-F was built as an interim 

facility in 1953 – 1954. Tritium extraction operations began in October 1955 after the reactor and 

separations start-ups. In July 1957, a larger tritium facility began operation in 232-H. In August 

1958, the capacity of 232-H was doubled. Also, in 1957 a new task was assigned to SRS – the 

loading of tritium into “reservoirs” that would be actual components of thermonuclear weapons. In 

October 1958, the F-Area Tritium Facility was shut down permanently (WSRC-IM-94-39). 

In addition to tritium exposures during the tritium recovery process in the 200-H Area facilities, there 

was the potential for tritium exposure to workers in the reactor buildings from leaks from the primary 

system. The reactors used “heavy water” (high in the deuterium isotope of hydrogen) as the 

moderator, which resulted in the production of tritium by neutron absorption in the deuterium isotope. 

Thus, there was the potential for tritium exposure to workers in the reactor buildings from leaks from 

the primary system. Construction workers would have been exposed during maintenance operations. 

Personnel Monitoring for Tritium 

NIOSH has obtained tritium bioassay data for workers starting in 1953. In 1979, the Health 

Protection Annual Radiation Exposure History Database (HPAREH) was developed, which recorded 

annual doses to workers (shallow, deep, neutron, and tritium). Information regarding the reporting 

levels and detection limits at SRS is included in the SRS TBD. Information in the bioassay logbooks 

show the highest pre-HPAREH results for the applicable monitoring years. 

Airborne Monitoring for Tritium 

Tritium exposure was controlled through a combination of facility design, training, personnel 

protective equipment, detection and measurement, and work controls (DPSPU-58-030-014; DPSPU62-30-5; DPSPU-62-030-025A). Tritium air concentrations were continuously monitored by stack 

monitors and by portable instruments. Three instruments were typically used because they met most 

requirements: a Kanne chamber (WSRC-IM-94-39), a portable tritium monitor (“sniffer”), and a small 

ionization chamber for taking grab samples. 

Controls for tritium in air are interrelated with bioassay and annual dose limits. For control purposes, 

a 20 μCi/L value has been used; personnel who assimilate tritium above this level are removed from 

all work where further uptakes of tritium may occur (DPSPU-62-030-025A). When the assimilated 

tritium exceeds 40 μCi/L, the employee is removed from all work related to radiation. In either event, 

the employee cannot return for further work in tritium atmospheres until the body burden is 10 μCi/L 

or below. Good control has been realized by requiring personnel working in tritiated atmospheres to 

wear plastic suits for line breaks or when airborne concentration exceeds 20E-5 μCi/cc. 

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Table 7-2 shows the tritium exposure experience at SRS for the years 1954 through 1961. 

Table 7-2: SRS Tritium Exposure Experience 

Year 1954 1955 1956 1957 1958 1959 1960 1961 

No. Assayed 2190 2310 2230 4000 3600 2590 2720 2690 

Tritium Assimilations 

0.37 - 1.0 millicurie * * * * 307 156 452 401 

>1.0 millicurie 0 2 13 49 33 22 79 73 

Max. Assimilation, millicurie - 1.2 153 8.8 3.6 3.2 3.9 14.6 

Source: DPSPU 62-30-5 

* Data not tabulated prior to 1958. 

Feasibility of Bounding Tritium Doses 

Construction workers exposed to tritium were monitored by urinalysis; therefore, personal bioassay 

data are available and reported by the site in their individual monitoring records. These data can be 

used to bound the individual worker’s dose from tritium. For workers who were not monitored, or 

whose records provided by the Department of Energy may be incomplete (e.g., missing data), tritium 

dose may be assigned based on information in Section 4.4.3 of the TBD (ORAUT-TKBS-0003). That 

is, based on their job category (ORAUT-OTIB-0014), internal doses of construction workers may be 

bound based on the tritium urinalysis reporting levels. 

7.1.1.2 Uranium 

The production of desired materials from the SRS nuclear production reactors required a combination 

of fissionable material to provide the neutrons (or the “fuel”) and fertile material to serve as targets. 

The basic fuel used in the SRS reactors was U-235. Basic targets consisted of U-238, which would be 

turned into plutonium. The uranium fuel or targets were manufactured at SRS in the Manufacturing 

Area (also known as M Area or 300 Area). The bare uranium slugs were typically made at Feed 

Materials Production Center and shipped to SRS for canning in the Fuel Slug Manufacturing Building 

(SRS at Fifty, 2002). The building was designed for the preparation of natural-uranium slugs and 

subsequent canning of the slugs in an aluminum sheath for protection from water corrosion. Based on 

the suitability testing of the material destined for the reactor (conducted at the M-Area test piles), the 

fuel and target assemblies advanced from solid slugs to tubular assemblies. SRS used nearly all of the 

Highly Enriched Uranium (HEU) from the recycled uranium from the INEEL in “driver” fuels, which 

in turn provided high flux. The first tubular production work at SRS was done in the Alloy Building 

in mid-1956, but this work was soon transferred to the Manufacturing Building, which was built in 

1956 and 1957 specifically for the tube manufacture. The recovery of uranium mixtures and enriched 

uranium occurred in the 200-F and 200-H Areas. H Area employed the “HM” process to recover 

enriched uranium. Many metric tons of uranium were used during the reactors’ operational periods. 

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Personnel Monitoring for Uranium 

Uranium bioassay has been conducted at SRS since the start of radiological operations. SRS technical 

documentation indicates that for earlier monitoring periods, the designations “enriched” and 

“depleted” reported for uranium analysis referred to analysis performed by alpha counting or chemical 

measurement (fluorometric method), respectively; these designations were not necessarily indicative 

of the degree of uranium enrichment. “EU” was the code used on employee bioassay cards for the 

gross alpha count method, and “DU” was used to designate the fluorophotometric method. Activity 

fractions for depleted, recycled, and highly-enriched uranium are listed in the TBD. Procedures for 

uranium urinalysis are adequately described in the TBD (ORAUT-TKBS-0003). Alpha spectroscopy 

has been used since 1990 for specific isotopic analysis. Information regarding the reporting levels and 

detection limits at SRS is included in the SRS TBD. 

A Progress Report issued in March 1961 (DPSP-61-1-2) states that analytical procedures require 

radioautographing on NTA slides to obtain the necessary sensitivity for most alpha emitters but that 

developments in low-level alpha counting indicates that the lengthy (seven days) radioautographing 

step may be replaced using silicon detectors and counters. 

Although not specifically mentioned in the bioassay records, analysis of U-233 would be conducted 

via the alpha-counting method with the same separation and measurement methodology used for other 

isotopes. However, no U-233 analytical results have been reported for SRS claimants. 

Feasibility of Bounding Uranium Doses 

Construction workers exposed to uranium were monitored by urinalysis; therefore, personal bioassay 

data are available and reported by the site in their individual monitoring records. These data can be 

used to bound the individual worker’s dose from uranium. For those workers who were exposed but 

might not have been monitored, sufficient data exists so that a co-worker model can be developed 

based on monitored workers for both enriched and natural/depleted uranium. Although many of the 

uranium bioassay logbooks are available, the co-worker model being developed will be based on the 

claimant data in NOCTS. The coding for these data was completed in late August 2008; the model is 

currently being developed and documented as of this writing. For the development of this model, 

there are 3796 fluorometric measurements of uranium from 1953 to 1992, 3133 alpha count 

measurements from 1955 through 1994, and 813 alpha spectroscopy results from 1994 through 2004. 

Due to the extensive bioassay monitoring records available to NIOSH, it is feasible to develop a coworker model; therefore, unmonitored doses can be bounded. 

7.1.1.3 Plutonium 

Plutonium, mainly Pu-239, was a major production material throughout the life of the SRS reactors. 

Pu-239 was created from U-238 by neutron bombardment in the reactors. When capturing a neutron, 

U-238 becomes an unstable U-239 isotope that decays to the more stable Pu-239 isotope by a two-step 

beta-negative decay process with Np-239 as the intermediate decay product (SRS at Fifty, 2002). The 

irradiated targets were transported to the 200-F Area and 200-H Area for product separation and 

packaging. The PUREX separation process was employed in F and H areas to recover plutonium, 

alternating process schedules for plutonium recovery. Wet chemical separations were accomplished 

in the canyon portion of the canyon building. Final products were solidified for shipment on the B46 of 94 
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Lines, and uranium was solidified for re-use or storage in the A-Lines (SRS at Fifty, 2002). Other 

minor isotopes of plutonium, (Pu-238, Pu-240, Pu-241) are by-products from the Pu-239 production 

processes. The production of Pu-239 began in 1953 and continued until 1989. By the early 1960s, the 

AEC had 14 production reactors - nine at Hanford and five at SRS. At the peak of production, these 

reactors produced around seven metric tons of plutonium per year. By 1964, the U.S. alone had over 

60 tons of plutonium. Between 1955 and 1964, SRS produced approximately one-third of the nation’s 

plutonium (DOE/DP-0137). 

During the 1950s and early 1960s, work was done by SRS to prepare radioisotope thermoelectric 

generators (RTGs) for the space program. The best heat source for the production of electricity in 

space was proven to be Pu-238. Production of Pu-238 began at SRS in the late 1950s and continued 

until 1986. During this period, Pu-238 was created by neutron irradiation of Np-237 targets in an 

increasingly-sophisticated series of reactor assemblies (SRS at Fifty, 2002). Plutonium recovery 

occurred in the Canyon Building and the B-Lines at the 200-F and 200-H Areas. Production 

quantities of Pu-238 were in the metric tons as SRS supplied the majority of the Pu-238 source 

material for the RTG heat sources for the space programs. 

Personnel Monitoring for Plutonium 

Urinalysis for plutonium was implemented in 1954. Samples were electroplated and activities were 

determined by gross alpha track analysis of exposed NTA emulsions. An analytical method to detect 

plutonium activity was implemented in 1959 that used ion exchange to separate plutonium; the 

minimum detectable activity was about the same as that of the alpha track method. Results were 

recorded as either Pu or as Pu238/Pu239. This method was used until 1964 when gross alpha analysis 

using a solid-state, surface-barrier alpha detector was used. In 1981, alpha spectrometry was 

implemented along with a new co-precipitation technique (although some with alpha track analysis 

continued until 1990). With the onset of this co-precipitation methodology, bioassay results for 

Pu-238 and Pu-239 were reported separately. Separation of plutonium with neptunium, actinides, 

uranium, and strontium from a single sample began in 2001. Alpha-emitting plutonium and neptunium 

isotopes were electrodeposited and counted by alpha spectrometry on a single planchet. Information 

regarding the reporting levels and detection limits at SRS is included in the SRS TBD. 

Feasibility of Bounding Plutonium Doses 

Construction workers exposed to plutonium were monitored by urinalysis; therefore, personal 

bioassay data are available and reported by the site in their individual monitoring records. These data 

can be used to bound the individual worker’s dose from plutonium isotopes. For those workers who 

were exposed, but might not have been monitored, a co-worker model can be developed based on 

monitored workers for Pu-238 and Pu-239. Although many of the plutonium bioassay logbooks are 

available, the co-worker model being developed will be based on the claimant data in NOCTS. The 

coding for these data was completed in late August 2008; the model is being developed and 

documented as of this writing. For the development of this model, there are 6186 results for total 

plutonium from 1954 to 1990, and 4076 results from 1981 through 2006 for the specific radionuclides 

Pu-238 and Pu-239. Due to the extensive bioassay monitoring records available to NIOSH, it is 

feasible to develop a co-worker model; therefore, unmonitored doses can be bounded. 

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7.1.1.4 Neptunium (Np-237) 

Neptunium-237 was the fertile material used for the production of Pu-238. Neptunium targets were 

manufactured in the 300 Area and were destined for the production reactors. Neptunium also was 

created when U-235 was irradiated. Therefore, Np-237 remained within the spent fuel matrices until 

they were processed in the Canyon Building in the 200-F and 200-H Areas and discharged as High 

Level Radioactive Wastes to the Waste Tanks. The production periods for Np-237 coincide with the 

reactors’ operational periods. However, quantities of Np-237 remain in the Waste Tanks until the 

present time. The amount of Np-237 handled at SRS was in the many-kilograms range. 

Personnel Monitoring for Neptunium 

SRS implemented urinalysis for Np-237 in 1959 (ORAUT-TKBS-0003). This method co-precipitated 

neptunium with calcium-magnesium ammonium phosphate, separated neptunium with ion exchange, 

followed by autoradiography on NTA emulsion for 10,000 minutes. While the MDA was not given, it 

was mostly likely similar to the MDA for plutonium alpha autoradiography method in use at the same 

time. The results were recorded as S.P. (special product). Information regarding the reporting levels 

and detection limits at SRS is included in the SRS TBD. 

In 1964, SRS implemented a new TIOA extraction and electro-deposition methodology. While the 

method was more time-efficient, the reported level remained the same as that of the previous method. 

Samples were analyzed by an off-site, commercial laboratory using extraction chromatography resin 

from 1993 through part of 1996. In 1996, analysis responsibility returned to SRS where alpha 

spectrometry was used for neptunium analysis. Since 2001, TEVA and TRU resins have used to 

separate neptunium from actinides and uranium, and alpha spectrometry has continued. Information 

regarding the reporting levels and detection limits at SRS, and the method for interpreting and 

applying values for dose reconstruction, are included in the SRS TBD. 

Feasibility of Bounding Neptunium Doses 

Construction workers exposed to neptunium were monitored by urinalysis; therefore, personal 

bioassay data are available and reported by the site in their individual monitoring records. These data 

can be used to bound the individual worker’s dose from neptunium. For those workers who were 

exposed but might not have been monitored, a co-worker model can be developed based on monitored 

workers for neptunium. Although the bioassay logbooks are available, the co-worker model being 

developed will be based on the claimant data in NOCTS. The coding for these data was completed in 

late August 2008; the model is being developed and documented as of this writing. For the 

development of this model, there are 304 measurements of neptunium from 1960 to 2004. Due to the 

bioassay monitoring records available to NIOSH, it is feasible to develop a co-worker model; 

therefore, unmonitored doses can be bounded. 

7.1.1.5 Americium (Am-241 and Am-243) 

Am-243 was one of the intermediate isotopes created in the progression of transforming Pu-239 to 

Cm-244 as well as one of the isotopes studied as a possible heat source. It was also used as target 

material for the production of transplutonium isotopes. Cm-244 is produced through successive 

neutron captures in Pu-239. The transmutation would take Pu-239 to Pu-240 to Pu-241, followed by 

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Pu-242, Pu-243, Am-243, Am-244, and finally Cm-244. A production campaign, Curium I, took the 

progression from Pu-239 to Pu-242. The targets from this irradiation were reprocessed and made into 

new targets for the second phase of this campaign, Curium II, which transmuted Pu-242 to the 

product, Cm-244, through Am-243 and Am-244 (SRS at Fifty, 2002). Because Curium II required a 

much higher neutron flux than ever been attempted, a demonstration program (called the High Flux 

Program) was implemented prior to the start of Curium II. In addition to supporting the curium 

program, the High Flux Program also demonstrated the applicability of high neutron flux for the 

generation of transplutonium isotopes. The target material for the High Flux Program consisted 

primarily of Pu-242. In addition, three 1-inch-diameter thimbles contained Am-243, Cu-244, and 150 

nuclides of 66 elements for nine universities and laboratories (WSRC-MS-2000-00061). Am-243 was 

generated during the irradiation processes in the High Flux Program and Curium II and was also used 

as targets for the High Flux Program. The first production period for Am-243 began in 1959 and 

continued through 1965. The production picked up in 1967 after a year of stoppage. The early SRS 

transplutonium programs generated 930 grams of Pu-242, 300 grams of Am-243, and 330 grams of 

Cm-244. By the end of the program, a total of 5.9 kilograms of curium and many kilograms of Am243 were generated (SRS at Fifty, 2002). Americium-241 is generated by the decay of Pu-241 and 

coexists in equilibrium with Pu-241 as its progeny. 

Personnel Monitoring for Americium 

Records of in vitro bioassay for trivalent actinides (americium, curium, and californium) showed 

urinalysis data back at least to about 1963. Bioassay for Am-243 was not needed before 1963. DPSP55-454-002 states: “Recovery of Pu242, Am243, and Cm244 – Fuel for the first transplutonium 

separation campaign consisted of 13 assemblies discharged from R-reactor in Apr 1963 after approx 

5.2 years’ irradiation.” Reporting levels for gross alpha determination of americium varied through 

1970. In 1971, gross alpha counting using a solid-state detector was implemented. SRS changed the 

procedure for radiochemical processing in 1990 Alpha spectrometry has been used since 1994. Up 

through the implementation of alpha spectroscopy in 1993, exposure location or other information 

concerning an intake was useful in the determination of the principal radionuclide since the results 

were obtained for trivalent actinides. Without sufficient information to separate activities to particular 

radionuclides, the entire activity was assigned to Am-241. Information regarding the reporting levels 

and detection limits at SRS is included in the SRS TBD. 

Feasibility of Bounding Americium Doses 

Construction workers exposed to americium were monitored by urinalysis; therefore, personal 

bioassay data are available and reported by the site in their individual monitoring records. These data 

can be used to bound the individual worker’s dose from americium. 

For those workers who were exposed but might not have been monitored, a co-worker model can be 

developed based on monitored workers for americium. Although many of the bioassay logbooks are 

available, the co-worker model being developed will be based on the claimant data in NOCTS. The 

coding for these data was completed in late August 2008; the model is being developed and 

documented as of this writing. For the development of this model, there are 926 trivalent 

measurements from 1963 to 1994, and 96 alpha spectroscopy measurements from 1993 through 2006. 

Due to the extensive bioassay monitoring records available to NIOSH, it is feasible to develop a 

co-worker model; therefore, unmonitored doses can be bounded. 

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7.1.1.6 Curium (Cm-244) 

In late 1963, Cm-244 was sought as a possible heat source for the thermoelectric generation of 

electricity, and it was also desired as a target material for the production of californium. Curium-244 

is produced through successive neutron captures in Pu-239. Gram quantities of Cm-244 were first 

produced in SRS reactors in 1962. In May 1964, the curium production campaigns (Curium I and II) 

began irradiation to produce kilogram quantities of Cm-244 for evaluation as an alternative to Pu-238 

and for use as target material to produce Cf-252. The targets from Curium I irradiation were 

reprocessed and made into new targets for the second phase of this program (Curium II) to complete 

the transmutation of Pu-239 to Cm-244. A separate High Flux Program or Demonstration was started 

in February 1965 and transmuted much of its Pu-242 targets to Cm-244. By 1967, a total of 5.9 

kilograms of Cm-244 had been made during the two curium production campaigns. Continued 

irradiation of plutonium target material eventually produced about 12 kilograms of Cm-244 (WSRCMS-2000-00061). 

Personnel Monitoring for Curium 

Records of in vitro bioassay for trivalent actinides (americium, curium, and californium) showed 

urinalysis data back at least to 1963. In 1971, gross alpha counting using a solid-state detector was 

implemented. SRS changed the procedure for radiochemical processing in 1990. Alpha spectrometry 

has been used since 1994. Up through the implementation of alpha spectroscopy in 1994, exposure 

location or other information concerning an intake was useful in the determination of the principal 

radionuclide since the results were obtained for trivalent actinides. Without sufficient information to 

separate activities to particular radionuclides, the entire activity can be assigned to Cm-244. 

Information regarding the reporting levels and detection limits at SRS is included in the SRS TBD. 

Feasibility of Bounding Curium Doses 

Construction workers exposed to curium were monitored by urinalysis; therefore, personal bioassay 

data are available and reported by the site in their individual monitoring records. These data can be 

used to bound the individual workers dose to curium. For those workers who were exposed but might 

not have been monitored, a co-worker model can be developed based on monitored workers for 

curium. Although many of the bioassay logbooks are available, the co-worker model being developed 

will be based on the claimant data in NOCTS. The coding for these data was completed in late 

August 2008; the model is being developed and documented as of this writing. For the development 

of this model, there are 926 trivalent measurements from 1963 to 1994, and 92 alpha spectroscopy 

measurements from 1994 through 2006. Due to the extensive bioassay monitoring records available 

to NIOSH, it is feasible to develop a co-worker model; therefore, unmonitored doses can be bounded. 

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7.1.1.7 Californium (Cf-252, including Cm-248 daughter) 

Cf-252 is one of the heaviest of the transplutonium elements and produces an intense neutron source 

from spontaneous fission. The program to specifically create californium (Californium I) began in 

August 1969 and continued until November 1970, with irradiation work conducted in K Reactor. The 

program generated the element in sufficient quantities to make the isotope available to industry and 

medical facilities. The production of Cf-252 entailed the addition of eight neutrons to the nucleus of 

the starting element, Cm-244, with the atomic number increasing from 96 to 98. The residues 

provided by Curium II were used as targets for the production of californium. When the program 

ended in November 1970, SRS had produced 2.1 grams, the largest amount of californium ever 

produced (SRS at Fifty, 2002). 

Personnel Monitoring for Californium 

Records of in vitro bioassay for trivalent actinides (americium, curium, and californium) showed 

urinalysis data back at least to about 1963. In 1971, gross alpha counting using a solid-state detector 

was implemented. SRS changed the procedure for radiochemical processing in 1990. Alpha 

spectrometry has been used since 1994. Up through the implementation of alpha spectroscopy in 

1994, exposure location or other information concerning an intake was useful in the determination of 

the principal radionuclide since the results were obtained for trivalent actinides. Without sufficient 

information to separate activities to particular radionuclides, the entire activity can be assigned to 

Cf-252. Information regarding the reporting levels and detection limits at SRS is included in the SRS 

TBD. 

Feasibility of Bounding Californium Doses 

Construction workers exposed to californium were monitored by urinalysis; therefore, personal 

bioassay data are available and reported by the site in their individual monitoring records. These data 

can be used to bound the individual workers dose to californium. For those workers who were 

exposed but might not have been monitored, a co-worker model can be developed based on monitored 

workers for californium. Although all of the bioassay logbooks are available, the co-worker model 

being developed will be based on the claimant data in NOCTS. The coding for these data was 

completed in late August 2008; the model is being developed and documented as of this writing. For 

the development of this model, there are 926 trivalent measurements from 1963 to 1994, and 92 

spectroscopy measurements from 1994 through 2004. With this bioassay monitoring records 

available to NIOSH, it is feasible to develop a co-worker model; therefore, unmonitored doses can be 

bounded. 

7.1.1.8 Thorium (Th-228, Th-232, and U-233) 

The details of the thorium processes and programmatic progress can be gleaned from the Monthly 

Progress Reports of the Works Technical Department beginning with the January 1953 edition and 

unfolding over subsequent years. The Feed Materials Production Center (FMPC) was the principal 

supplier of thorium to SRS. Much of that was canned by Sylvania Electric Products (SEP) Company 

with subsequent fuel fabrication with a small fraction at SRS. 

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In August 1954, Test Authorization Number 3-138, DPSOX-00530, AlSi Canning Thorium Slugs 

(Huntoon, 1954), in Building 313-M was approved by the Plant Manager, SRL Director and seven 

others, including the Health Physics Superintendant. By November 1954, four quatrefoil fuel 

assemblies had been produced and were included in the L-2 reactor loading for irradiation testing 

(DPSP 54-1-10; DPSP 54-1-11). The plan was to return the slugs to the hot labs of R or P areas when 

this test was completed (DPSP 54-1-11); in May 1955, four irradiated thorium slugs were returned to 

SRL for testing in the 100-R hot lab (DPSP 55-1-8-DV). 

Early in 1955, Sylvania Electric Products (SEP) was contracted to produce 11,000 hot-press bonded 

thorium slugs due to their demonstrated capabilities to produce uranium slugs using the same process 

and the low yields from the competing AlSi hot-dip process at SRS. A second order for 15,000 more 

canned thorium slugs was largely fulfilled in August 1955 (DPSP-55-454-3). The SEP thorium fuel 

was used for the R-7 full-core reactor loading (DPSP 55-1-8-DV). Programmatic changes ended the 

need for thorium fuel. In September, the SEP thorium program was reviewed, resulting in a plan to 

dispose of the unused inventory by stripping the cans and returning the metal to FMPC (DPSP 54-1­

11) 

In September 1956, SRS received a request to irradiate five tons (approximately 5000 slugs) of 

thorium to produce U-233 (DPSP 56-1-9). Shipping continued in August and a plan was presented to 

retain two irradiation thorium slugs from R-11 to make activity measurements at the test facility in 

105-K Disassembly Area during the following year to define the absolute decay schemes associated 

with the irradiated fuel (DPSP 57-1-8). This research began according to plan in November 1956 

(DPSP 57-1-11-DV). In December 1956, Building 313-M personnel were reduced to 109 wage roll 

and 21 supervisors from the previous level of 154 wage roll and 26 supervisors (DPSP-55-454-3). 

In December 1958, ORNL proposed that pellets of mixed 4% enriched UO2 and ThO2 be irradiated at 

SRS. It is not known if this occurred; however, in June 1959 a procedure was developed and tested 

for making a final arc weld on aluminum-canned hot-press bonded slugs containing powdered and 

compacted thorium oxide. 

In October 1961, irradiated metal waste, including thorium and uranium with 8400 Ci of fission 

products was shipped from the SRL high-level caves to Building 643-G. In 1962, Bridgeport Brass 

Company extruded 43,000 kg of thorium metal which was machined and canned at Sylcor. In 

February 1963, 6000 Mark VII-T canned thorium slugs were finished at SRS for use in the L-1 

charge. Then 10,000 more were finished in November 1963 for use in the L-2 charge. 

In 1964, vibratory compaction equipment was installed to produce thorium oxide (thoria) slugs. By 

August 1964, 13,600 slugs (each containing 3.6 kg of thoria) were being produced at a rate of 120 per 

shift. It was noted that contamination was controlled satisfactorily (DPSP-55-454-3). Also in 1964, 

Sylcor canned and delivered 9600 Mark VII-TS solid thorium slugs. In 1965, there were 28,160 

thorium oxide slugs canned (DPSP-55-454-3). In 1966, thoria slug fabrication facilities in Building 

313-M were improved in preparation for a new production campaign. In 1967, process development 

continued until March when operator training began (DPSP-55-454-3). During 1967, there were 

23,962 Mark 50A slugs produced (each containing 3.7 kg of thoria), and 17,098 Mark 50B slugs (each 

containing 1.7 kg of thoria). All but 2,100 of these were produced between April and August in threeshift operations. During November and December, the remaining 2100 slugs were produced on a 

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single-shift schedule (DPSP-55-454-3). In January and February 1968, an additional 3,760 thoria 

slugs were produced, completing the campaign (DPSP-55-454-3) 

Personnel Monitoring for Thorium 

Urine bioassays for thorium were conducted in accordance with a procedure (Ellett, 1958), and work 

technical report (DPSP 56-1-8). A re-sampling threshold was established at SRS as the bioassay 

program matured (Logbook P006, p. 78). A 1956 bioassay logbook dedicated to thorium bioassays 

shows that 224 bioassay were performed on 168 different individuals. The bioassay program was 

carefully done, as evidenced by the 42 blanks and 45 spikes recorded in the logbook. With minor 

exceptions, the net sample results were similar to the sample blanks. At least one additional thorium 

bioassay result predating the logbook has been found in a claimant's dosimetry record, suggesting that 

more samples will be discovered when earlier logbooks are examined. No urine bioassay samples for 

thorium have been found after 1956. This date roughly coincides with the end of bare thorium metal 

canning at SRS, which may provide an explanation. Contemporary ICRP biokinetics models for 

thorium show that urine bioassay lacks the sensitivity to demonstrate compliance with regulatory 

limits, and therefore, urine bioassay is not used as part of a routine thorium bioassay program. The 

lack of information on thorium biokinetics was mentioned in the procedure (Ellett, 1958). The 

Internal Dosimetry Technical Basis Manual (WSRC-IM-90-139, Rev. 1) identifies bioassay methods 

for Th-228 and Th-232 which would have been performed after 1960. 

In-vivo counting was implemented in 1960 and would have been sensitive to thorium daughter 

products. Whole body counting and chest counting were routinely used as part of the internal 

dosimetry program at SRS. 

Airborne Monitoring for Thorium 

Representative air sampling data have been requested for locations where un-encapsulated thorium 

was processed prior to 1960. ICRP 75, General Principles for the Radiation Protection of Workers, 

discusses the uses of area air sampling to establish trends and representative air sampling to quantify 

the extent of airborne contamination to which workers are likely to be exposed. The air sampling 

requirement specified in the August 1954 Test Authorization Number 3-138 suggests that the air 

sampling data will satisfy the "representative sampling" definition in ICRP 75 which states that 

representative sampling typically involves using fixed samplers at locations intended to be reasonably 

representative of the breathing zone of the workers. If air sample data are available as expected, then 

intake rates can be calculated based on an assumption of a default breathing rate and an appropriate 

occupancy factor in these areas using the air sample data. 

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Feasibility of Bounding Thorium Doses 

Some construction workers exposed to thorium were monitored by urinalysis in 1956 and other 

workers were monitored by in vivo analysis after 1960. Using the in vivo data, doses received from 

potential exposures to thorium starting in 1960 can be bounded for production workers, and therefore, 

for construction and maintenance trade workers by a combination of: 

•	 in vivo counting data that became available after 1960; and 

•	 representative air sampling data from locations where un-encapsulated thorium was processed 

prior to 1961. 

NIOSH is reviewing the 1956 bioassay data to determine the adequacy of that data for assigning 

internal dose. At this point, NIOSH has not determined whether the reconstruction of thorium doses 

received prior to 1960 is feasible. A follow-up to this report will be issued once the air sample data 

prior to 1960 has been reviewed. 

7.1.1.9 Fission and Activation Products 

Europium-152/154 was produced primarily as fission products in the fuel and also by neutron 

activation of reactor control rods. Europium, along with other fission products, was separated in the 

200-F and 200-H Areas from the product streams containing uranium and plutonium. Because Eu-152 

had useful application as a food sterilizer and as a source for radiation therapy, SRS isolated Eu-152 

from the Purex waste stream for evaluation in a 1967 campaign. It has been estimated that kilograms 

of Eu-152 were produced at SRS. 

In SRS reactors, a small percentage of the uranium in the fuel was converted into other fission 

products as by-products of energy generation. Activation products resulted from neutron capture by 

other materials, such as structural components of the reactor, the reactor coolant, control rods, or fuel 

and target claddings and containers. The dose-significant fission and activation products at SRS 

include Mn-54, Ce-144, Zn-65, Sr-90, Ru-106, Sb-125, and Cs-134/137. Fission and activation 

products remain in the fuels and control rods until they are reprocessed and separated from uranium 

and plutonium at the Canyon Building in the 200-F and 200-H Areas and discharged to the Waste 

Tanks as high level radioactive wastes. Production of these radionuclides coincides with the reactors’ 

operational periods. However, metric tons of these radionuclides currently remain in the Waste 

Tanks. 

Personnel Monitoring for Fission and Activation Products 

Initially, intakes of gamma-emitting fission products were monitored by urinalysis, which began in 

1954 when the reactors went on line. The start date for urinalysis for strontium radioisotopes was not 

determined more specifically than the “late 1950s.” However, Boni discusses a bioassay procedure in 

use at the time for rapid estimation of the total amount activity of beta-gamma-emitting isotopes in a 

urine specimen with recoveries of Sr-89, Sr-90/Y-90, Zr-95/Ni-95, Ce-144/Pr-144, Fe-59, Cr-51, and 

Zn-65 greater than 90% and Co-60 of 85% (Boni, 1959). Fission product bioassay was performed to 

determine intakes of Tm-170. 

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Welders and grinders at completion of their work in the (reactor RDZ) area, but prior to leaving the 

area, were required to submit urine samples for induced activities (750 mL), in addition to tritium 

samples as of August 26, 1957. The sample was to be started soon enough for completion prior to 

leaving the area (e.g., approximately ten days before) (Health Physics, 1957). 

Gamma-emitting fission products were monitored by urinalysis prior to whole body counts, but also 

into the mid-1960s. Strontium isotopes (and potentially cerium and promethium) were separated by 

alkaline earth phosphate co-precipitation followed by counting on a GM counter. This procedure was 

in use from start-up through 1969 when replaced by liquid ion exchange methods that separated the 

yttrium progeny and beta proportional counting. Since 2001, strontium was separated as part of the 

use of TRU and TEVA resins for plutonium, uranium, actinides, and strontium (ORAUT-TKBS0003). The procedure separated and counted radionuclides of Sr, Y, Ce, and Pm. Analyses for beta 

and for gamma-emitting fission and activation products have been referred to as FPIA (Fission 

Product Induced Activity). 

Whole-body counting has been performed since 1960 when the SRS Whole Body Counting Facility 

was completed. A March 16, 1961 report (DPSP-61-1-2) stated that the body burdens for 84% of 37 

Separations employees in the 200-F Area, measured with the whole body counter, had detectable 

amounts of Ce/Pr-144, Ru/Rh-106, or Zr/Nb-95. The maximum amounts found were 21 nanocuries 

Ce/Pr-144, 32 nanocuries Ru/Rh-106, and 1.8 nanocuries Zr/Nb-95. Minimum detectable activities 

for in vivo analysis of fission and activation products are given in ORAUT-TKBS-0003 (pp. 72-73). 

Airborne Monitoring for Fission and Activation Products 

SRS performed air sampling through 1956 to determine if respiratory protection should be used. The 

requirement to wear a respirator was based on the measured level of Fe-59. Due to multiple slug 

failures during 1957, experimental and developmental programs in disassembly areas, and interchange 

of disassembly equipment between areas, the masking requirements basis was changed to unknown 

fission products. 

Feasibility of Bounding Fission and Activation Product Doses 

Construction workers exposed to fission and activation products were monitored by either urinalysis 

or whole-body counting or both; therefore, personal bioassay data are available and reported by the 

site in their individual monitoring records. These data can be used to bound the individual worker’s 

dose from fission and activation products. For those workers who were exposed but might not have 

been monitored, a co-worker model can be developed for fission and activation products based on 

monitored workers. Although bioassay logbooks are available, the co-worker model being developed 

will be based on the claimant data in NOCTS. The coding for these data was completed in late 

August 2008; the model, which will provide intakes of the radionuclide providing the highest organ 

dose, is being developed and documented as of this writing. With more than 4,700 results in NOCTS, 

it is feasible to develop a co-worker model; therefore, unmonitored doses received from exposures to 

fission and activation products can be bounded. 

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7.1.1.10 Cobalt-60 (Co-60) 

Co-60 was produced by the irradiation of natural cobalt (Co-59) in the SRS reactors. Cobalt-60 emits 

very penetrating gamma radiation. Sources made of Co-60 can be used for cancer treatment, food 

sterilization, radiography, and as heat sources for thermoelectric generators. Small amounts of 

low-specific-activity Co-60 (~50 curies per gram) were produced in SRS reactors as early as 1955. 

Beginning in 1965, the High Flux Program (see the Cm-244 description), provided an opportunity to 

make large quantities of very-high-specific activity Co-60 when Co-59 was the preferred material for 

control and safety rods in high flux reactor cores (WSRC-MS-2000-00061). The first production 

program (1955-1958) made Co-60 with a specific activity of 50 Ci/g. The second period of 

production (1959-1964) created Co-60 with 100 Ci/g. The third program (1964-1967) generated 

material with 700 Ci/g. The cobalt produced at SRS is estimated to be in the kilogram range (SRS at 

Fifty, 2002). 

Personnel Monitoring for Cobalt 

Exposures to Co-60 have been monitored using in vivo analysis (whole body counting) since about 

1960. Prior to that, and continuing through the mid- to late-1980s, gamma-emitting fission products 

were monitored by urinalysis using both gross beta analysis and gamma-spectrometry analysis. 

NIOSH has obtained the logbooks for many urinalysis results. See the discussion on fission/activation 

product bioassay under Section 7.1.1.9. 

Feasibility of Bounding Cobalt Doses 

Construction workers exposed to cobalt were monitored by urinalysis and/or whole body counting; 

therefore, personal bioassay data are available and reported by the site in their individual monitoring 

records. These data can be used to bound the individual worker’s dose from Co-60. For those 

workers who were exposed but might not have been monitored, a co-worker model can be developed 

based on monitored workers for cobalt. Although all of the bioassay logbooks are available, the coworker model being developed will be based on the claimant data in NOCTS. The coding for these 

data was completed in late August 2008; the model is being developed and documented as of this 

writing. For the development of this model, more than 4700 measurements from 1954 to 2004 for 

fission product determination are available. Due to the extensive bioassay monitoring records 

available to NIOSH, it is feasible to develop a co-worker model; therefore, unmonitored doses can be 

bounded. 

7.1.1.11 Polonium 

Starting in June 1966, Po-210 was produced at SRS by the irradiation of bismuth slugs as part of the 

Curium II project (DPSP-67-1281). SRS produced 324,200 curies from June 1966 through 1967. The 

slugs were shipped to Mound in Ohio. Recovery of polonium took place at Mound, not at SRS. 

From 1967 through 1970, SRS performed research using laboratory and plant-scale equipment to test 

the engineering feasibility of using liquid metal distillation to separate up to 10 kg of Po-210/year 

from irradiated bismuth (DP-1222). Long-lived (>60 day half-life) gamma-emitting impurities had to 

be separated from the Po-210 distillation. While the percentage-by-weight of those impurities was 

small, SRS felt that they could result in unacceptable levels of gamma radiation. Some separation was 

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achieved by alloying for decay. A high-purity grade of bismuth was irradiated and then cooled for 

about 15 months. After the cool-down, Zn-65 was distilled with Po-210, while some Sb-124 and 

Co-60 activity remained in the bismuth residues. 

Personnel Monitoring for Polonium 

As noted above, the research conducted in 773-A was a very small scale. NIOSH has obtained 

bioassay logbooks for analysis of Po-210 and believes the individuals involved in the project were 

monitored via bioassay. NIOSH has requested the bioassay procedure used to interpret the results. 

Together, the data and procedure will be used to bound internal doses received from exposure to 

polonium for these few workers. 

Feasibility of Bounding Polonium Doses 

NIOSH can bound potential internal exposures received from polonium using the available bioassay 

data. NIOSH will use this limited data set to develop a co-worker intake model for use in bounding 

doses. For construction workers who may not have been monitored, internal exposures received from 

potential intakes of polonium can be bounded through the application of the co-worker model. 

7.1.2 Internal Dose Reconstruction Feasibility Conclusion 

The guidance of ORAUT-OTIB-0052 is to determine internal dose of construction workers using the 

same method applied to all other SRS workers. 

The dose to SRS construction workers who were monitored for internally-deposited occupationallyrelated radionuclides may be bounded with sufficient accuracy based on their reported bioassay 

monitoring results. 

Additionally, workers who were not monitored, or whose records provided by the Department of 

Energy may be incomplete (e.g., missing data), may be assigned internal doses based on information 

in Section 4.4.3 of the SRS TBD (ORAUT-TKBS-0003) and from intakes derived in co-worker 

models discussed in Section 7.1 of this Evaluation Report. 

Doses received from potential exposures to thorium starting in 1960 can be bounded using the data 

and methods discussed in Section 7.2.1.5. The feasibility of bounding radiation doses received from 

exposure to thorium from start-up through 1959 is being investigated. That determination is reserved. 

7.2 External Radiological Exposure 

ATTRIBUTION: Section 7.2 and its related subsections were completed by Mike Mahathy, Oak Ridge 

Associated Universities; Eugene Potter, Mel Chew and Associates, Inc; Bryce Rich, Mel Chew and 

Associates, Inc.; and Darin Hekkala, Dade-Moeller, Inc. These 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. 

SRS was built to produce the materials used in the fabrication of nuclear weapons, primarily tritium 

and Pu-239. Five reactors were built to produce nuclear materials through irradiation of target 

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materials with neutrons. Support facilities included two chemical separations plants, a heavy water 

extraction plant, a nuclear fuel and target fabrication facility, and waste management facilities. HEU 

fuels were used within the reactor core to create higher neutron flux areas, which in turn allowed more 

efficient production of other special isotopes (i.e. Pu-238, Cf-252, and others). In addition, the 

Savannah River Laboratory conducted research in support of the production operations and special 

isotope production campaigns. In the following subsections, the source term, personnel monitoring 

methods, and feasibility of reconstructing photon and neutron doses are discussed. 

7.2.1 Photon Exposures 

7.2.1.1 Source Term 

The photons associated with each of the three main steps of material production are discussed below 

along with exposures experienced at the Savannah River Laboratory. The three main steps in nuclear 

material production include the fuel and target fabrication, reactor operations, and chemical 

separations. 

300 Area - Fuel and Target Fabrication 

The primary target materials loaded into the SRS reactors consisted of depleted uranium for Pu-239 

production. These targets were manufactured in the fuel and target fabrication area (300 Area). The 

photon energy spectra from uranium covers both intermediate and high-energy photons; both are 

easily measured using film badges and TLDs. The external photon dose rate from this source was 

relatively low so work around the targets before irradiation did not require any special shielding and 

workers could generally handle the fuel and targets wearing gloves. During special isotope 

production campaigns, the photon energy spectra and dose rates could be significantly different 

compared to the standard depleted uranium work, with increased low-energy photons for some target 

materials to which the film badge responded with significant increase in indicated dose response. For 

these campaigns, it is evident from the dosimetry records and the monthly works technical reports that 

the SRS Health Physics Group was aware of these differences and implemented a special method for 

interpretation of the darkening on the badge film by using a low-photon energy calibration curve to 

account for the difference in photon energy spectra. 

100 Area – Reactor Operations 

At the reactor areas, the fuels and targets were loaded into assemblies and then placed in the reactor. 

The reactor was then operated for a set time period depending on the isotopic production goal of the 

irradiation. During operation, biological shields around reactor core reduced the dose levels from 

prompt fission photons and fission product photons from irradiated fuel. The top of the reactors were 

at grade level and access to the lower levels was restricted/controlled during operation. After the fuel 

and targets were irradiated in the reactors, they were removed and allowed to decay in the spent fuel 

pool or disassembly basin area of the reactor. The transfer of the fuel and targets from the reactor to 

the disassembly area was conducted remotely because the mixed fission products and activation 

products made these targets highly radioactive so that shielding was necessary to protect workers, 

minimize exposures, and maintain levels below regulatory limits at maximum. Once in the 

disassembly area, the spent fuel and targets were maintained in a pool of water that had the dual 

function of cooling the thermally-hot targets and shielding workers from photon radiation. While the 

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photon emissions associated with these irradiated fuels and targets covered the whole energy 

spectrum, the predominate photon energies were greater than 250 keV. This energy range is 

considered “high energy” and is relatively easy to measure using film dosimeters and TLDs. With 

very few exceptions, all workers exposed to these photons were monitored for external exposure. 

200 Area – Chemical Separations (Canyons) Operations 

The irradiated fuels and targets were sent to the chemical separations areas via shielded containers on 

railcars for nuclear material (product) extraction. Once in the canyons, the irradiated targets were 

dissolved and materials of interest were separated. The uranium was recovered to be “recycled” in the 

A-line; the plutonium was recovered in the B-Line; and the remaining mixed fission and activation 

products went to the waste tanks (Tank Farms). As discussed above, uranium is relatively evenly split 

between intermediate and high-energy photons and the highly-radioactive waste predominately 

consisted of high-energy photons. Photons associated with recovered plutonium are generally of low 

energy; however, when quantities of Pu are concentrated in a workplace, moderate dose rates can 

result. The photon dose rate from plutonium is not nearly as high as the waste material, but is 

significantly higher than uranium due to higher specific activity and increased neutron emission rates. 

700 Area – Savannah River Laboratory 

The photon exposures and associated energy spectra at the Savannah River Laboratory was complex 

and changed over time due to multiple operations, special isotope campaigns, and other research that 

was conducted. However, the film dosimeters and TLDs were capable of measuring the full photon 

energy spectra (low, intermediate, and high). 

For more information on photon sources and the energy spectra associated with SRS operations, see 

ORAUT-TKBS-0003, Table 5.3.4.1-1. 

7.2.1.2 Personnel Monitoring Data 

SRS had an extensive radiation safety monitoring program to measure and control external photon 

radiation exposure in the workplace. Measurements were conducted using portable radiation 

instruments, pocket ionization chambers, and personnel dosimeters. Control of external radiation 

hazards was established through the implementation of Special Work Permits (SWPs) and the 

establishment of radiation control areas. The predominant method of personnel radiation exposure 

monitoring was the individual dosimeter. The first beta/gamma dosimeters used at the onset of 

operations in November 1951 were provided by the Oak Ridge National Laboratory (ORNL). ORNL 

provided beta/photon film dosimeters and neutron nuclear track, type A (NTA) emulsion dosimeters 

until SRS implemented an in-house dosimetry program in March 1953 (WSRC-RP-95-234). SRS 

dosimetry methods evolved over the years as improved technology was developed and the complex 

radiation fields were better understood. The initial personnel dosimeter program consisted of a 

two-element film dosimeter capable of measuring low, intermediate and high energy photons; 

however, the dosimeter over-responded to low and intermediate energy photons (Wright, 1958). 

The early (pre-1958) photon dose measurements for SRS employees were hand-entered onto cards 

and filed in individual employee folders. Each SRS employee has a folder that contains this early 

monitoring data. For some construction trades workers, however, the monitoring data were recorded 

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on 3x5 visitor cards and cycle monitoring log sheets that are filed by time and by area. During this 

evaluation, these visitor cards and cycle monitoring data were located and requested from the site. 

They are undergoing classification review as of the writing of this report. When a claim is filed by an 

SRS worker, in addition to searching their regular records, the SRS Radiation Records group also 

searches the visitor cards for the individual worker. NIOSH is currently verifying that the information 

on the monitoring sheets matches the visitor cards and that these are received for each monitored 

claimant. This verification is expected to be completed in February 2009. At this time, NIOSH does 

not have any reason to believe that these monitoring records are different from the information 

included in each individual claim record. 

In November 1959, a new multi-element film dosimeter and automatic film processor was introduced 

(WSRC-RP-95-234, Davis 1960 DP-471). The use of the multi-element filters improved the low and 

intermediate energy response (i.e., they better corrected for the lower-energy over-response). The 

data from this dosimeter as well as the data from the last year from the previous dosimeter (1958) 

were entered into a new IBM computer system for dose record tracking. On a quarterly basis, SRS 

produced a quarterly printout of all monitored workers. In addition to the standard printout, a special 

printout of the people who exceeded the administrative threshold for the quarter, and for those who 

were projected to exceed the annual limit of 5 rem per year (i.e., for the second quarter the annual total 

was increased to 2.5 rem, and for the third quarter the annual total was increased to 3.75 rem.) During 

this evaluation, NOSH has obtained a copy of all of the quarterly dosimetry printouts between 1958 

and 1988. 

In April 1970, TLDs replaced film badge dosimeters at SRS. This style dosimeter has been, and still 

is, the standard for external dose monitoring. The record-keeping system (IBM Quarterly Printouts) 

remained the same. In 1979, the Health Protection Annual Radiation Exposure History (HPAREH) 

system was initiated. At that time, all of the current employees’ annual radiation exposure records 

were hand-entered into this new electronic system. The subsequent annual doses were then 

transferred each year to HPAREH. This continued until the initiation of a fully-electronic system in 

April 1989 called the Health Protection Radiation Exposure Database (HPRED). 

As indicated above, individual radiation dose records from personnel dosimeters worn by workers and 

co-workers are available from 1951 to present. Figure 6.1 illustrated that on the order of 6000 

workers per year were routinely monitored for radiation exposure. The SRS Health Physics 

department required dosimeters be issued to all workers who entered radiation controlled areas. 

During NIOSH review of SRS documentation, an exception to this monitoring practice was noted in a 

Test Authorization (DPSOX-254) from November 1953 to July 1955. In this authorization, only 

persons assigned to a particular location were assigned film badges (i.e., visitors were not always 

assigned badges). The restriction on this authorization is that only work with natural uranium was 

allowed and the practice was only allowed until dosimeters were available at the area gate house. This 

Test Authorization was a stopgap between the start-up period and the regular issuance of film badges 

from the area gate houses at the access point to the specific area. The Test Authorization was 

rescinded in July 1955 because it was no longer needed. 

Another type of monitoring exception noted in the SRS Radiation Survey Log Sheets (RSLS) is that 

for some construction work, access controls were put in place to limit the exposure received by 

construction workers during large modifications. The access controls effectively cordoned off the 

work zone within a larger radiation control area. Within this area, radioactive materials were removed 

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from the work zone so that the only external exposure would originate from outside the work zone. 

The perimeter of the work zone was then monitored through the use of survey instruments and/or 

dosimeters hung on the perimeter towards the radioactive materials to ensure that radiation levels 

within the zone were below regulatory requirements. As noted, these exceptions are documented in 

the Radiation Survey Log Sheets (RSLS). 

Thus, while the normal practice was to monitor all individuals within a radiation control area, there 

were occasions, as indicated by workers during the outreach meetings, in which it would appear as if 

the workers were not monitored within a radiation area. However, based on NIOSH’s review, area 

monitoring was being conducted to ensure that exposures were less than regulatory requirements. 

Since this control was in effect, the use of a co-worker model for workers within these controlled 

zones supports NIOSH’s ability to bound the photon dose for the class under evaluation. 

7.2.1.3 Feasibility of Estimating or Bounding Photon Dose 

The SEC evaluation effort was focused on the ability to bound dose for the class (unmonitored 

construction trade workers). At SRS, construction trades workers (CTW) were generally monitored 

for radiation exposure just like other production workers; when they entered a radiological controlled 

area, they were issued a film badge dosimeter (or TLD in later years). However, there are a few 

occasions where a construction worker might not have been monitored and radioactive materials could 

have been present (e.g., 1953-1955). However, as discussed in the preceding sections, these doses can 

be bounded using co-worker monitoring data. 

Dose reconstructions are performed for unmonitored CTW at SRS using co-worker dose data 

provided in ORAUT-OTIB-0032. The application of co-worker data for all dose reconstructions (for 

unmonitored periods) is described in ORAUT-OTIB-0020. Guidance in ORAUT-OTIB-0052 is 

applied in concert with ORAUT-0TIB-0020 for unmonitored CTW. ORAUT-OTIB-0052 

recommends that a correction factor of 1.4 be applied to co-worker doses used when reconstructing 

doses for unmonitored CTW. In other words, the dose is increased 40% over the dose that would be 

applied to an unmonitored non-CTW determined to have been routinely exposed to radiation. 

As noted previously in Section 6.2, the co-worker annual dose data provided in ORAUT-OTIB-0032 

uses the HPAREH database, which is known to be incomplete for years prior to 1979. However, 

NIOSH has reviewed the HPAREH database and compared it to a sample of the full hardcopy data 

from the Quarterly Dosimeter Reports and found that the dose distributions in HPAREH may be used 

to bound the general population’s annual photon dose and, therefore, can be used to bound individual 

worker photon exposures. 

7.2.2 Neutron Exposures 

Neutron exposures at SRS were limited to specific operations within several areas. The sources of 

neutron exposure are shown in Table 7-3 and are limited to areas such as reactor operation and any 

process with significant quantities of transuranics (TRU) (i.e., with spontaneous fission and α/n 

reaction potential) or other high neutron-emitting sources. The 200 and 700 areas were the most 

important with regard to neutron exposures. While one might generally consider neutron exposures to 

be significant around the production reactors, the exposure potential was actually quite low due to 

extensive shielding. Another area with intermittent neutron exposure was the fuel and target 

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fabrication area (300 area). The exposures in this area depended on the campaign being conducted. 

In the subsections below, the sources of neutrons, the personnel and area monitoring that was 

conducted, and the feasibility of reconstructing neutron doses are presented. 

Table 7-3: Sources of Neutrons 

Source Dates Exposure Potential Comments 

Five Production R 12/53 – 6/64 Operating reactors extensively surveyed during start-up – 

Reactors P 2/54 – 8/88 

K 10/54 – 7/92 

L 7/54 – 6/88 

C 3/55 – 6/85 

essentially no neutron exposure potential. Monitoring for 

neutrons was performed in preparation for occasional 

entries during operation to equipment inside lessershielded cubicles. Neutron personnel dosimetry was 

provided. Typically 1:10 n/p ratios were measured. 

F Separation Plant F 11/54 – 1/57 TRU isotopes were separated and processed. Neutrons 

plus A, B, and JB 2/59 – 1989 were produced through spontaneous fission and α/n 

product lines; H 7/55 – 3/59 reactions. Moderated spectra of varying degrees were 

H Separation Plant 5/59 – 1989 encountered. Facilities were shut down for remodeling 

plus HB product lines and process change during the early years. Surveys and 

and NSR Facility personnel neutron dosimetry were provided, as needed. 

R&D, Special 

Product, Technology, 

Vault & Storage 

Facilities, 

uranium target and 

fuels fabrication 

PuFF and PEF 

1978-1989 

773-A, 313M, 320M, 

321M, 322M, 341M 

1952 - 1989 

Quantities of TRU processed and/or stored 

773-A: Cf-252 product 

300 Areas: Primarily uranium target and fuels fabrication 

and development 

Analytical and A-772-F and 772-1F Smaller quantities of TRU and variety of neutron sources 

Calibration Facilities production control 

labs. 736-A 

Calibration Facility 

1953 - 2003 

in the Calibration Facility, PuBe, PuF4 and Cf-252 

7.2.2.1 Source Term 

300 Area - Fuel and Target Fabrication 

For certain campaigns, such as the production of trans-plutonium isotopes (californium and curium), 

plutonium-aluminum (Pu-Al) targets were manufactured in the 300 area. In addition, for the 

production of Pu-238, neptunium-aluminum (Np-237-Al) alloy targets were manufactured. The 

neutron dose rate from these targets was low to moderate, with the plutonium being much higher than 

neptunium from a neutron dose rate standpoint. 

During this evaluation, NIOSH found that routine neutron measurements of these targets were 

conducted and recorded on the Radiation Survey Log Sheets (RSLS) and occasionally reported in the 

Health Physics Monthly Reports. Table 7-4 presents a few examples of the neutron dose rate 

measurements on these targets recorded in the 300/700 Area monthly reports. 

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Table 7-4: Typical Neutron Dose Rates on Special Targets 

Target Date Distance and # 

tubes 

Fast Neutron 

Dose Rate 

(mrem/hr) 

Total Dose 

Rate 

(mrem/hr) 

Nf to Photon 

Ratio 

Pu239-Al January 1971 18” 1-Tube 4 200 0.02 

Pu239-Al January 1971 18” 6-tubes 15 800 0.02 

Pu242-Al June 1969 18” 1-Tube 1 2 1.0 

Pu242-Al June 1969 18” 6-tubes 1 6 0.2 

Np237-Al January 1971 18” 1-Tube 1 80 0.01 

Np237-Al January 1971 18” 6-tubes 2 300 0.01 

For the Pu-239-Al and Np-237-Al targets, the vast majority of the external dose came from the 

photons not neutrons. However, the Pu-242-Al targets illustrated a significantly higher neutron-tophoton ratio. 

100 Area – Reactor Operations 

As a general rule, neutron doses around the production reactors occurred only while the reactor was 

operating, with the exception of neutron source handling. As noted in the photon section, the top of 

the reactors were level with the ground floor; thus, most of the neutron exposure was below-grade. 

The areas below grade were controlled/restricted during reactor operations. The reactors’ biological 

shielding prevented most neutrons from reaching the general operations areas occupied by workers. 

From NIOSH’s review of the Radiation Survey Log Sheets, there appears to be a couple of areas that 

personnel were occasionally allowed to enter during reactor operations that could have resulted in 

neutron exposure. These were the crane wash area and the labyrinth to the pump room. However, as 

noted in the Radiation Survey Log Sheets, neutron survey measurements were recorded prior to entry 

and neutron dosimeters were required. Based on a limited review of the Radiation Survey Log Sheets, 

the general N:P ratio appears to be about 1:10, or 1 mrem of neutrons per 10 mrem of photons. 

200 Area - Separations 

At the 200 Area plutonium processing facilities, neutron radiation is generated from plutonium or 

other transuranics by spontaneous fission and/or by alpha particle interaction with light elements such 

as oxygen, fluorine, and carbon in the product processing lines, which have lesser shielding than the 

separations canyons. The predominate neutron exposures occurred along the B-lines or plutonium 

process lines. The three B-lines were the FB Line (early operations), JB Line (later operations), and 

the HB line. Neutrons were produced through spontaneous fission and α/n reactions. Moderated 

spectra of varying degrees were encountered at locations along these lines in which there were 

quantities of TRU/TRP. Throughout the operations, facilities were occasionally shut down for 

remodeling and process change; thus, neutron monitoring can appear to be intermittent. However, 

from the review conducted for this evaluation, the monitoring appears to correspond well with 

production operations. 

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700 Area – Savannah River Laboratory 

The predominate neutron exposures at the Savannah River Laboratory were associated with research 

being conducted on various plutonium and trans-plutonium isotopes and special materials 

processing/production. The production of Cf-252 is one of the most notable neutron exposures within 

this area as californium is a strong neutron emitter due to its relatively high probability (cross-section) 

for spontaneous fission. 

7.2.2.2 Neutron Personnel Monitoring Data 

Personnel neutron dose monitoring was provided through the use of dosimeters for workers in areas in 

which survey data indicated a neutron dose rate greater than 1 mrem/hr, or where the potential for 

neutron dose rates existed. Neutron monitoring using NTA film began in August 1953 and was 

replaced by TLND in June 1971, which in turn was replaced by the automated Panasonic UD-809 

albedo neutron dosimeter in January 1995. CR-39 track-etch detector type dosimeters were 

experimentally investigated and researched, but never used to assign personnel exposures. 

The number of workers monitored for neutron dose varied with time due to changing radiological 

conditions, changing standards and regulations, and material production. Because of the restricted 

areas of the plants in which detectable neutron radiation was present, relatively few workers were 

monitored for neutron exposure; fewer still members of the CTW class currently under evaluation, 

since members of this class were generally not routinely assigned work in operating areas in the 

reactors, TRU production areas, or experimental/R&D areas where significant quantities of neutrongenerating materials or neutron sources were present during operational periods. 

Figure 7-2 depicts the number of beta-gamma badges and Figure 7-3 depicts the number of NTA 

badges processed per cycle from the Personnel Meters Reports for the 200-F Area. This area was 

chosen as an example because this area was one of the highest neutron-exposure areas at SRS. At the 

time of this writing, the data in these graphs are incomplete due to delays in receiving the Personnel 

Meters Reports from SRS (i.e., data from 1960 and part of 1962 are missing). The interesting point of 

these figures is that while the number of photon dosimeter badges remained relatively constant from 

1958 through 1970, the number of neutron dosimeter badges increased from 1961, in which less than 

2% of the 200-F area workers were monitored for neutron exposure, to near 15% in 1969. In 1970, 

SRS began to monitor some workers on a quarterly basis while others (more highly-exposed) 

remained on a monthly basis. Considering this component, the relative fraction of the most 

highly-exposed workers monitored for neutrons increases from 15% to near 25%. 

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Figure 7-2: Number of Beta/Gamma (Photon) Film Dosimeter Badges Processed per Cycle for the 200-F Area. 

Figure 7-3: Number of Neutron Film Dosimeter Badges Processed per Cycle for the 200-F Area. 

During this evaluation, NIOSH located and reviewed all of the NTA data for SRS from 1961 through 

1971 when the TLND was implemented. These data are recorded on the NTA Process and Inventory 

Log Sheets and contain the number of tracks read per field of view and the conversion/calibration 

factor for each batch of NTA film used in an area. The inventory sheets are organized by area and 

contain individual monitoring results as well as summary results of the positive neutron doses. At the 

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time of this writing, NIOSH has requested these records and they are currently undergoing 

classification review. NIOSH expects to receive these records by the end of November 2008. 

In the past, NTA film data has been considered to be of limited value due to the relatively poor 

response (under-response) of the film to intermediate and low energy neutrons, generally below about 

500 keV. This deficiency has been recognized industry-wide and the subject of continuing R&D with 

a continuing series of corrections applied to recorded NTA results. SRS, in addressing the neutron 

measurement technological difficulties, participated with other AEC/DOE laboratories and facilities in 

the solution of the issue both at SRS and industry-wide. As evidence of this continuing participation, 

an extensive study was performed and published that included complete neutron spectral 

measurements on all SRS workstation locations at which neutron exposures could occur. This study 

is available and can be used as another means of validation of the establishment of any necessary 

correction factors for the early NTA dose measurements (PNL-6301). However, the response or 

degree of under-estimation is highly dependent on the calibration source used. At some other 

facilities, a relatively low neutron energy source such as PuF4 was used so that the under-response was 

generally less than a factor of two. However, at SRS the calibration source was initially a PoBe 

source and later changed to a PuBe source, both of which are high-energy neutron sources. As 

indicated in Table 6-3, prior to the change from NTA film to TLND, a comparison was conducted at 

the Plutonium Finishing Area. This comparison indicated that the NTA film under-responded by an 

average of 3.9. Thus, to accurately bound an individual worker’s neutron dose at SRS, the NTAmeasured neutron dose should be multiplied by 3.9. At the time the SRS TBD was written, the actual 

correction factor was not known because the results from this historical comparison study had not yet 

been located. As a result, a neutron-to-photon ratio was developed using TLND data (post-1971). 

This N:P ratio was back-extrapolated to cover the NTA time period. Now that more specific 

information related to the under-response is known, the current N:P ratio in the SRS TBD can be 

validated against the corrected NTA dose values as developed, using the recently recovered 

information. 

7.2.2.3 Feasibility of Bounding Neutron Doses 

At present, NIOSH has not found any evidence that the N:P ratios used in the SRS TBD are 

unreasonable, and therefore, cannot be used for dose reconstruction. However, given the new data 

that has been found as a result of this evaluation, NIOSH now has the ability to either validate or 

correct the current neutron-to-photon ratio used for SRS dose reconstructions. 

For this validation, an N:P ratio will be developed based upon paired NTA and photon measurement 

data post-1962. The distribution of this ratio will be compared to the TBD value and NIOSH will 

either validate the current listed TBD N:P ratios or update them. The completion date for this 

validation is not known as of this writing because the information NIOSH reviewed in August 2008 

has not yet been cleared and received. Upon receipt of this information, NIOSH expects the 

evaluation will take 3-4 months to complete. The NTA data in the claimant files will be corrected to 

total neutron dose from all energies based on the data in DP-MS-69-4 and the additional evaluations. 

Since personnel neutron monitoring data are limited prior to 1962, paired neutron and photon dose 

rate measurements will be extracted from the Radiation Survey Log Sheets and used to develop the 

N:P ratio. This survey-based N:P ratio will be compared to both the current value in the TBD as well 

as the value developed using the corrected NTA monitoring data. The timeline for completion of this 

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component is significantly longer than the NTA analysis because the individual survey forms that 

contain the paired measurements must first be tagged for scanning. The sheets must then be coded 

and the analysis conducted. For the analysis conducted at the Hanford Site, this effort took 

approximately six months after the initial data review/capture. Due to NIOSH and DOE resource 

limitations, NIOSH’s tentative schedule is to review/capture this information in February 2009; thus, 

the validation using the survey-based data is expected to be complete in August 2009. 

As noted during this evaluation, NIOSH did not uncover any evidence to indicate that the N:P ratios 

in the SRS TBD are not bounding. Through the limited reviews of the Radiation Survey Log Sheets, 

the N:P ratios appear to fall within the range specified in the TBD. NIOSH recognizes that further 

evaluation is now warranted given the extensive monitoring records now available. 

As for the feasibility of neutron dose reconstruction, if neutron dose records exist for the individual 

worker, these should be used to reconstruct the neutron dose. For early time periods (NTA era), the 

measured neutron dose should be multiplied by a factor of 3.9. If neutron monitoring records do not 

exist, and it is established that the worker was potentially-exposed to neutrons in accordance with 

OCAS-TIB-007, then the neutron dose is calculated using the N:P ratios presented in the SRS TBD. 

The use of the N:P ratio is generally applied to pre-1971 data; however, for some workers, the ratio is 

also applicable post-1971. Effectively, for unmonitored construction trades workers, the co-worker 

photon dose multiplied by 1.4 (ORAUT-OTIB-0052) constitutes the base photon dose from which the 

N:P ratio is then applied to reconstruct the neutron dose. 

In summary, the current neutron dose reconstruction methodology is considered to be sufficiently 

accurate for bounding neutron doses at SRS. During this evaluation, NIOSH has discovered 

additional neutron monitoring data that can be used to validate the current methodology or to revise 

the methodology, as appropriate. Regardless of whether the evaluation of the new data validates the 

current method or results in a new method, NIOSH believes that the extensive amount of data 

discovered is sufficient to support the establishment of a bounding neutron dose for the class under 

evaluation. 

7.2.3 External Dose Reconstruction Feasibility Conclusion 

The available monitoring data for photon and neutron doses, and the process and source term 

documentation, have been examined and found to be adequate for estimating bounding external doses 

for the class. The analysis of newly-acquired photon dose data not previously included in the 

HPAREH database shows that the current annual co-worker dose values are appropriate for dose 

reconstruction when individual monitoring results are not available. Neutron doses, for years when 

NTA film was used, can be reconstructed by adjusting (correcting) the NTA film dosimetry results or 

by applying the current neutron-to-photon ratio, or by developing a new neutron-to-photon ratio based 

either on NTA film data and/or on neutron radiation survey data. Therefore, the external dose to SRS 

construction workers may be bounded with sufficient accuracy based on the reported monitoring 

results and the assessment of the external monitoring results, as described in this section. 

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7.3 Evaluation of Petition Basis for SEC-00103 

The following subsections evaluate the assertions made on behalf of and specific to petition 

SEC-00103 for the Savannah River Site. 

7.3.1 Dose Reconstructions for Construction Workers 

SEC-00103, Attachment 1, Page 43 of 91: Construction Workers have missing external monitoring 

data, for all or parts of their employment periods, or “zero” results recorded for an entire year even 

though working in ”radiation areas” or have been involved in radiation exposure incidents. 

This issue was addressed and resolved in Section 6.2. Based on the review of the annual dose 

distributions, and specific evaluation of the construction trades workers, NIOSH concludes that the 

HPAREH database used in ORAU-OTIB-0032 is sufficiently robust and can, therefore, be used to 

bound external photon and neutron doses. 

7.3.2 Consideration of Construction Workers’ Involvement in Incidents 

SEC-00103, Attachment 2, Page 64 of 91: NIOSH has not considered the involvement of construction 

workers in radiation incidents or accidents. 

NIOSH has considered each of these declared situations. In reviewing the SRS Site Incident database, 

NIOSH found that construction workers were recorded as being involved in about 25 percent of 

incidents. NIOSH further found that SRS HP provided monitoring for the particular situations (e.g., 

surface contamination monitoring, personnel contamination monitoring, follow-up contamination 

monitoring, personnel nasal smears, bioassay, and documentation for each involved worker). More 

discussion is presented in Section 5.2.3. 

7.3.3 Exposure Patterns of Construction Workers 

SEC-00103, Attachment 2, Page 64 of 91: In regards to the SRS Site Profile: The Site Profile is 

skewed toward production workers who work in one facility, or area, for a long time and does not 

account for the very different exposure patterns of construction workers. 

Members of the proposed SEC Class were employees of DuPont and of private contractor 

organizations and were provided the same Health and Safety support as other operations and 

maintenance workers. When any worker entered a radiologically-controlled work area, SRS required 

dosimetry, protective clothing, and equipment as required for the job, and bioassay as needed. Based 

on the review described in Sections 5.2.3, 6.1, and 6.2, NIOSH concludes that the HPAREH database, 

used in conjunction ORAU-OTIB-0032, ORAU-OTIB-0052, and internal intake co-worker data, is 

sufficiently robust and can, therefore, be used to bound external and internal doses for the class under 

evaluation. 

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7.3.4 1990 Tiger Team Assessment 

SEC-00103, Attachment 2, Page 64 of 91: NIOSH has not considered deficiencies in the radiation 

monitoring programs such as those identified by the 1990 Tiger Team (DOE/EH-0133). 

NIOSH has reviewed the 1990 Tiger Team report as part of this SEC Evaluation Report. The Tiger 

Team identified 94 findings pertaining to radiation protection. A numerical summary of these items 

by category is shown in Table 7-5. 

Table 7-5: Numerical Summary of Tiger Team Radiation Protections Findings 

Finding Category No. of Findings 

Contamination Control 7 

Radioactive Material Control 11 

Inadequate Air Sampling / Monitoring 9 

Effluent Environmental Monitoring 4 

Instrument & Calibration 22 

General Practice / Organization 11 

ALARA 9 

Training 9 

Procedures & RWP 7 

Dosimetry 5 

Total 94 

NIOSH reviewed the five findings pertaining to dosimetry: 

• Truck Drivers Wearing Dosimeter on Chest (source was behind driver) 

NIOSH Response: While a posterior to anterior geometry can be used to account for dose, the 

external dose can be bounded using Section 7.2. A technical information bulletin will be 

published for use individual dose reconstructions to correct for the geometry. 

• TLD Storage Racks Inside RCA (entry violation) 

NIOSH Response: In the case that dosimeters were stored inside an RCS, and workers were 

required to enter the RCA in order to get their dosimeter, this would result in a dosimeter-recorded 

reading higher than the actual radiation dose received by the individual. Therefore, this issue does 

not impact NIOSH’s ability to bound the dose for the class under evaluation. 

• Inadequate Monitoring of Casual Visitors Inside RCA (they were not performing work) 

NIOSH Response: Any person entering an RCA was required to wear a dosimeter, and if they 

were not, they were in direct violation of procedures. All visitors entering an RCA were issued 

visitor dosimeters. Based on the review and assessment performed by NIOSH for this report, this 

issue does not impact NIOSH’s ability to bound the dose for the class under evaluation. 

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• Workers Observed in RCA Not Wearing Proper Dosimetry Devices 

NIOSH Response: Any person entering an RCA was required to wear a dosimeter, and if they 

were not, they were in direct violation of procedures. All visitors entering an RCA were issued 

visitor dosimeters. Based on the review and assessment performed by NIOSH for this report, this 

issue does not impact NIOSH’s ability to bound the dose for the class under evaluation. 

• Inadequate Follow-up on Delinquent Bioassay 

NIOSH Response: NIOSH is developing a co-worker model based on the claimant data in NOCTS 

to account for intakes of workers who may not have been monitored during periods of 

employment. A separate evaluation has been completed that establishes the principle that, under 

certain conditions, the data in NOCTS are representative of a site’s population generally 

(ORAUT-OTIB-0075). Based on the review and assessment performed by NIOSH for this report, 

this issue does not impact NIOSH’s ability to bound the dose for the class under evaluation. 

7.3.5 Work in Non-radiological Areas 

SEC-00103, Affidavits, Pages 8, 23, 24, 25, and 27 of 91: Several affidavits claim instances of 

workers who were working in supposedly non-radiological areas, only to later find out that the area 

was contaminated and/or a radiation area. 

As noted in Section 7.3.2, NIOSH found in the SRS Site Incident database that construction workers 

were recorded as being involved in about 25% of incidents. NIOSH further found that SRS HP 

provided monitoring for the particular situations (e.g., surface contamination monitoring, personnel 

contamination monitoring, follow-up contamination monitoring, personnel nasal smears, and 

bioassay. In some instances, low levels of radiation were detected in the roped-off area but were not 

high enough to require personnel dosimetry according to standing procedure (i.e., anticipated 

exposures would be below permissible levels. From this analysis, NIOSH has established that areas 

where radiological work was conducted were monitored and that construction workers working in 

those areas were monitored. In the event it is determined that a specific worker was not monitored 

during a specific job, missed dose can be bounded using criteria provided in Sections 7.1 and 7.2. 

Based on the review and assessment performed by NIOSH for this report, this issue does not impact 

NIOSH’s ability to bound the dose for the class under evaluation. 

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7.3.6 Cover-up of Incidents 

SEC-00103, Affidavits, Page 15 of 91: Statements were provided that alleged radiological incidents 

were covered up. 

NIOSH has found no information to substantiate that radiological incidents were covered up, or to 

substantiate that some were not. It is at least plausible to expect that some radiological 

incidents/accidents were not formally reported due to lack of radiological impact. In the example 

cited in the petition, the individual involved indicated that subsequent to the incident, he had left 

bioassay samples. As a result, due to the routine monitoring that was conducted, the potential for a 

serious overexposure to have gone unnoticed even with a cover-up is considered to be relatively small. 

Even if the potential was larger, NIOSH finds that bound doses derived using both routine personnel 

monitoring data available in NOCTS and HPAREH, and co-worker intakes (Section 7.1) would 

include doses received during potential “covered-up” incidents. 

7.4 Other Potential SEC Issues Relevant to the Petition Identified During the 

Evaluation 

ATTRIBUTION: Section 7.4 and its related subsections were completed by Ed Scalsky, Dade-Moeller, 

Inc.; Mike Mahathy, Oak Ridge Associated Universities; Eugene Potter, Mel Chew and Associates, 

Inc. These 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. 

During the feasibility evaluation for SEC-00103, a number of issues were identified that needed 

further analysis and resolution. The issues and their current status are discussed in the following 

subsections. 

7.4.1 Radionuclides from Special Campaigns Not Included in the TBD 

Potential intakes of construction workers from exposures to, and subsequent intakes of, radionuclides 

handled or generated during Isotope Production Campaigns were not assessed. 

NIOSH reviewed bioassay logbooks, comparing the dates given for analyses of specified 

radionuclides to the dates of SRS isotope campaigns. Results of that comparison are provided in 

Figure 7-1. Monitoring for exposure to all isotopes was performed except for curium-244 from 1959 

through 1963. SRS Health Physics procedures for radiological monitoring applied to “All 

Departments and Construction” (DPSOP-40). Methods of reconstructing potential internal doses are 

discussed in Section 7.1 

7.4.2 Tank Farm Exposures 

During evaluation of the proposed class, NIOSH interviewed several former workers. A former 

worker stated that construction workers had performed frequent work at the Tank Farms, that the 

work was not monitored and that the source term had not been characterized. 

NIOSH evaluated the source term of the tank farms, which consisted of tritium, plutonium, uranium 

(including enriched [1.1%] uranium), neptunium, americium, curium, thorium, polonium, and fission 

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products. The activities ranged from high-level to low-level. From the examination of the SRS Site 

Incident database, NIOSH found that the Tank Farms were staffed by Health Physics and that there 

are monitoring data available for SRS incidents. Furthermore, NIOSH has established that SRS 

radiological workers were monitored for external and internal radiation exposure. For workers who 

should have been monitored but were not, both external and internal doses can be reconstructed using 

co-worker distributions for beta-gamma- emitting radionuclides, applying external correction factors 

given in ORAUT-OTIB-52 and applying intake correction factors established for construction 

workers based on stay times and ORAUT-OTIB-52 parameters. An internal co-worker intake 

distribution will be developed using all DOE-supplied bioassay data stored for SRS claimants in 

NOCTS. NIOSH has demonstrated that bioassay data reported for claimants of any particular DOE 

sites is representative of that site’s population of bioassay data (ORAUT-OTIB-0075). 

7.4.3 Forestry Workers 

During the same interviews mentioned in Section 7.4.2, a former Forestry Service employee stated he 

worked at SRS where he tested and helped clear trees in the canyons and other areas; his office was 

in Building 760G. He stated that he had not been monitored. 

NIOSH found that the worker was not monitored for external radiation exposure during his stay at 

SRS and employment with the Forestry Service. The worker did receive a whole-body count in 1992 

while working as an escort. While the worker would not be a member of the SEC class under 

evaluation, NIOSH finds that doses to forestry workers can be bounded using environmental internal 

doses, and dose conversion factors given in ORAUT-TKBS-0003, pp. 188-192. Intakes were 

calculated with the assumption that all foodstuffs were grown in the areas of highest radionuclide 

concentrations and that the radionuclide content in the plant mass can be calculated by use of transfer 

factors given in PNNL-13421. The transfer factors provide a ratio of the plant mass to the soil 

concentration (dry mass-to-dry mass). External doses can be bounded using assumptions and 

environment dose conversion factors given in ORAUT-TKBS-0003, pp. 164-187. 

7.4.4 Special Tritium Compounds 

Does internal uptake of special tritium compounds result in a longer bodily retention of tritium than 

other forms of tritium? 

NIOSH states that this issue does not prevent such doses from being bounded due to the large amount 

of tritium results available for SRS workers. The method for evaluating dose associated with special 

tritium compounds is documented in ORAUT-OTIB-0066, Calculation of Dose from Intakes of 

Special Tritium Compounds. This OTIB, coupled with the available SRS tritium monitoring data (as 

discussed in Section 7.0), supports NIOSH’s ability to bound the dose associated with special tritium 

compounds for the class under evaluation. 

7.4.5 Bounding of Doses from Intake of Thorium 

Can potential internal doses received from exposure to thorium be bounded for construction workers 

with sufficient accuracy? 

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NIOSH has analyzed its ability to bound internal thorium dose for the class under evaluation. The 

results of that analysis are presented in Section 7.1.1.9. NIOSH concludes that, for the period from 

1960 through 2003, available personnel monitoring data and dose reconstruction methods support the 

ability of NIOSH to bound the thorium internal dose for the class. However, NIOSH is continuing to 

evaluate the period from start-up through 1959, with that conclusion being RESERVED. 

7.4.6 Completeness of the HPAREH Database 

The Advisory Board contractor (SC&A) provided NIOSH with an inventory of supplemental records 

that may be beneficial in dose reconstruction. These records are not always provided in the claimant 

file submitted for dose reconstruction. No effort has been made to evaluate the completeness of the 

HPAREH file used in the development of the external co-worker model. The integrity of the HPAREH 

file for use in co-worker modeling is questionable given the absence of much of the data for workers 

terminating employment prior to 1979. 

NIOSH has evaluated a comparison of doses in the Quarterly Summary Reports with those in 

HPAREH (see Section 6.0 of this report). As discussed, not all historical monitoring data was stored 

in the HPAREH database. The evaluation of the comparison was performed to examine the effect of 

the missing data on NIOSH’s ability to bound the dose for the class under evaluation. NIOSH’s 

review shows that the distribution of doses for construction workers in HPAREH is somewhat higher 

than the distribution of doses using Quarterly Summaries for most years. However, the construction 

worker Quarterly Summary distribution is bounded in all evaluated cases when compared to the 

HPAREH distribution for all monitored workers (which is the method applied in ORAUT-OTIB-0052 

and the SRS external co-worker study). Based on the assessment of this approach, further refinement 

of the HPAREH distribution based on collection of additional data would only result in the reduction 

of the bounding values for the class. However, for the years where only a small percentage of 

HPAREH data exist (pre-1958 time frame), NIOSH will continue analysis of visitor logs and cards 

and refinements of the process to confirm this approach. 

Based on the availability of SRS documentation, and NIOSH’s review of the hardcopy records in 

comparison to the HAPREH database, there does not appear to be any significant administrative 

practice or data issue that would affect the integrity of the recorded SRS doses. The results of 

NIOSH’s evaluation on this matter demonstrate that the use of HPAREH data is bounding for 

construction worker external photon doses. 

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7.4.7 Dosimetry Uncertainty 

SC&A raised concerns about the beta/gamma correction factors. For the purpose of this evaluation, 

NIOSH states that only the SEC issues related to bounding the dose for the class must be considered 

in this report. As such, the evaluation of TBD correction factors or adjustments to recorded dose 

(which do not inhibit the ability of NIOSH to bound external doses) are generally considered Site 

Profile/TBD issues and not SEC issues (although some topics may be addressed in this evaluation for 

the purpose of completeness in the dosimetry accuracy discussions). In the interest of completeness, 

NIOSH is responding to the following uncertainty issues raised by SC&A. 

1.	 Calibration of dosimeters at 0 degrees is often not representative of incident angles encountered 

in the field and result in an underestimation of the true exposure that is being measured. 

An evaluation of several variables impacting the accuracy of the SRS Film Badge System was 

documented in a report prepared by Wright (Wright, 1958). This evaluation includes measured 

directional response characteristics of the SRS beta/photon film dosimeter throughout a full 360°

rotation. Figure 7-4 (reproduced from the Wright report shows iso-exposure circles at 0.9, 0.8 and 0.7 

in relation to the open-window exposure at 1.0 (i.e., 100%). The response to radium photon radiation 

by the dosimeter’s shielded region (dashed lines) is effectively equivalent to 1.0 based on the 

calibration for a 150-degree arc for an anterior irradiation. Overall, the badge responded better than 

0.9 and 0.8 over 90 % and 98 % of the viewing angles, respectively. Based on these figures, exposure 

from the posterior would result in an overresponse of the shielded region (i.e., greater than 1.0) 

In Figure 7-4, the notable reduction in response shown at 0° or 180° for an anterior irradiation is 

minimized in the workplace because of movement of the dosimeter while being worn by the worker, 

and by the broad radiation fields characteristic of typical workplaces due to scatter and large and/or 

multiple sources of radiation. The response of the dosimeter’s open-window region (solid line) is 

similar to the shielded response. 

Effects of energy dependence on the directional response are presented in the International Agency for 

Research on Cancer (IARC) Intercomparison Study (Thierry, 2002) of ten widely-used dosimetry 

systems throughout the world. The IARC Study illustrated that film dosimeters typically overestimate 

the delivered dose for photon radiation greater than about 80 keV in rotational, isotropic, and 

anterior-posterior exposure geometries relative to the typical method of calibration using radium in an 

anterior-posterior geometry. 

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Figure 7-4: Directional Response of Savannah River Plant Film Badge (Gamma) (Wright, 1958) 

2. The on-phantom correction factor of 1.119 may be too low for photon energies between 30 and 

250 keV. 

Reporting on the work of Goodwyn, Taylor provided the correction factors that were applied to bring 

the SRS dosimetry calibration into closer agreement with other sites and accepted standards (WSRCRP-95-234). For the period prior to 1986, for all beta/photon dosimeters, a correction factor of 1.119 

is applied; for 1986, a correction factor of 1.039 is applied; for the years 1987 to the present, no 

correction factor is required. 

The original work by Goodwyn has been requested from SRS and is undergoing classification review. 

As reported in WSRC-RP-95-234, SRS staff provided in the Goodwyn reference an evaluation of the 

overall adjustments necessary to arrive at a consistent estimate of the deep dose equivalent (Hp(10)) 

as employed in dosimeter performance studies at that time using ANSI N13.11-1983. This apparently 

represented their judgment of complex parameters, as noted in the SRS TBD. SRS staff had 

knowledge of follow-up dose evaluations for incidents where workplace directional and spectral 

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characteristics were considered, and comparisons of the dosimeter- and instrument-measured dose. 

The SRS TBD team concluded that (for dose reconstructions) an increase in the assigned dose by a 

factor of 1.119 should be employed in addition to other dose reconstruction considerations described 

in OCAS-IG-0001. 

3. The TBDs generic standard deviation value of 30% is likely to be low for film dosimeters prior to 

1971. Early film dosimeters are likely to have a workplace standard deviation of at least 40%. 

Uncertainty in the dosimeter results is an important component of external dose reconstruction. A 

systematic personnel meter (beta-gamma film dosimeter) audit record was routinely produced (SRS, 

1959). Visitor dosimeters were exposed to known doses of beta, gamma, and in some instances, 

X-ray radiation ranging from 0 to 300 mrem. The dosimeters were read and the results reported to 

Health Physics department management. A summary statement at the end of each report shows the 

average error for dose from gamma radiation, beta radiation, and sometimes the dose from beta plus 

gamma radiation. The April 3, 1961 audit marked the first use of the automatic densitometer in this 

testing program. Figure 7-5 shows the gamma dose results of this testing program based for the 

period March 19, 1959 through September 18, 1963. Additional records have been requested that 

may document results before and after this period. In general, the measured error seldom exceeded 

±20%. The dosimeters tended to be biased high and the beta dose errors exceeded gamma dose errors. 

All of the measured errors tended to be smaller after introduction of the automatic densitometer. 

Figure 7-5: SRS Film Dosimeter Audit Results for March 1959 through September 1963 

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7.4.8 Early Construction Worker Monitoring Data 

The TBD does not address the consistency of the SRS internal and external monitoring program for 

different operations and through time. This is especially important in the case of early workers who 

may not have had routine monitoring commensurate with their exposure. 

NIOSH finds that the HPAREH data can be used to support bounding external doses for the class 

under evaluation. NIOSH can continue to use the external monitoring data recorded at SRS to 

develop and further analyze the statistical relationship between HPAREH and the SRS-supplied data. 

NIOSH has assessed the available bioassay data (in NOCTS personnel records and SRS logbooks) 

and is establishing an internal co-worker study based on the NOCTS data (to be supplemented by 

other bioassay data, as necessary). Based on NIOSH’s reviews, the monitoring and radiological 

program information and data support that all workers entering a radiological area at the site were 

appropriately monitored (DPSOP-40). The only exception to this may have been when special 

exclusion zones were established, and when this occurred, the outer boundaries of the radiological 

areas were monitored. 

7.4.9 Early Construction Worker Neutron Monitoring Data 

NIOSH identified a lack of neutron monitoring data for all workers including construction workers in 

the time period from 1954 through 1960. Personnel monitoring for neutrons was only required when 

workers entered a radiation area where the neutron dose rate exceeded 1 mrem/hr. Thus, some 

unmonitored exposure to neutrons was likely. From the review of the SRS Incident database, NIOSH 

has determined that construction workers worked in areas of potential neutron exposures. 

NIOSH has reviewed SRS radiation survey reports to determine if workplace neutron measurements 

were conducted in parallel with photon survey measurements. NIOSH has verified that this parallel 

logging did occur. Paired measurements were recorded on individual Radiation Survey Log Sheets 

and are available in the SRS records holdings. The combination of the corrected NTA data, along 

with this survey data, will be used by NIOSH to establish the neutron-to-photon ratio for use in 

bounding neutron doses from 1954 through 1960. This methodology is discussed in more detail in 

Section 7.2. 

7.4.10 Neutron to Photon Ratio 

SC&A stated that both the TLND recorded neutron doses between 1971 and 1995, as well as the 

pre-1971 neutron doses (derived by neutron-to-photon ratios) suffer from a high degree of 

uncertainty. 

NIOSH has addressed this issue; see Sections 6.2, 7.2.2.1, 7.2.2.2 and 7.2.2.3 for discussion. 

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7.5 Summary of Feasibility Findings for Petition SEC-00103 

This report evaluates the feasibility for completing dose reconstructions for employees at the 

Savannah River Site from January 1, 1950 through December 31, 2007. NIOSH found that the 

available monitoring records, process descriptions, and source term data available are sufficient to 

complete dose reconstructions for the class under evaluation. 

Table 7-6 summarizes the results of the feasibility findings at SRS for each exposure source during 

the time period January 1, 1950 through December 31, 2007. 

Table 7-6: Summary of Feasibility Findings for SEC-00103 

January 1, 1950 through December 31, 2007 

Source of Exposure Reconstruction Feasible Reconstruction Not Feasible 

Internal1

 X 

 - U X 

 - Pu X 

 - Am X 

 - Np X 

 - Cm X 

 - H-3 X 

 - Cf X 

- Th X (1953-1959 decision reserved)

 - Fission Products X 

External X

 - Gamma X 

 - Beta X 

 - Neutron X 

 - Occupational Medical X-ray X 

1

 Internal includes an evaluation of urinalysis (in vitro), airborne dust, and lung (in vivo) data 

As of October 1, 2008, a total of 1798 claims have been submitted to NIOSH for individuals who 

worked in construction at the Savannah River Site. Dose reconstructions have been completed for 

1358 individuals (~75%). 

8.0 Evaluation of Health Endangerment for Petition SEC-00103 

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 bound 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 

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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. 

After examination and cross-comparison of the available monitoring data, source term information, 

and process information, NIOSH’s evaluation determined that it is feasible to bound radiation dose for 

members of the NIOSH-evaluated class with sufficient accuracy based on the sum of information 

available from available resources. Modification of the class definition regarding health 

endangerment and minimum required employment periods, therefore, is not required. 

9.0 Class Conclusion for Petition SEC-00103 

Based on its full research of the class under evaluation, NIOSH found no part of said class for which it 

cannot bound radiation doses with sufficient accuracy. This class includes all construction workers 

who worked in any area at the Savannah River Site during the period January 1, 1950 through 

December 31, 2007. However, NIOSH has reserved the feasibility determination for thorium 

exposures from January 1, 1950 through December 31, 1959; NIOSH is continuing to evaluate the 

thorium bounding approach for this time period. 

NIOSH has carefully reviewed all material sent in 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 (SRDB), for 

information relevant to SEC-00103. 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 contends that it has complied with these 

standards of performance in determining the feasibility or infeasibility of reconstructing dose for the 

class under evaluation. 

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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: 

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 

DNA-1251-1-EX, Compilation of Local Fallout Data from Test Detonations 1945-1962 Extracted 

from DASA 1251, Volume 1, H. A. Hawthorne; May 1, 1979; SRDB Ref ID: 8859 

DOE/DP-0137, Plutonium: The First 50 Years – United States Plutonium Production, Acquisition, 

and Utilization from 1944-1994, February 1996; SRDB Ref ID: 12292 

DOE/EA-0948, Environmental Assessment Operation of the HB-Line Facility and Frame Waste 

Recovery Process for Production of PU-238 Oxide at the Savannah River Site, U.S. Department of 

Energy, Savannah River Operations Office; April 1995; SRDB Ref ID: 48942 

DP-1222, Separation of 210Po from Bismuth by Liquid Metal Distillation, R. F. Bradley, L. F. 

Landon, G. W. Gibson, E. I. du Pont de Nemours and Company, June 1970; SRDB Ref ID: 48630 

DP-MS-69-4, A Thermoluminescent Personnel Neutron Dosimeter, A. Korba, J. E. Hoy, E. I. du Pont 

de Nemours and Company, October, 9, 1969; SRDB Ref ID: 48550 

DPSOP-40, Savannah River Plant Radiation and Contamination Control – History Copy, Vol. I and 

II; E. I. du Pont de Nemours and Company Explosives Department, date unknown; SRDB Ref ID: Not 

yet in SRDB 

DPSOP-47, Westinghouse Savannah River Company Radiobioassay Laboratory Procedures, Rev. 14, 

SRP Health Protection Department; July 1987; SRDB Ref ID: 45959 

DPSOX-254, Test Authorization 5-1: Film Badge Issuance, C. M. Patterson; November 23, 1953; 

SRDB Ref ID: 49202 

DPSP-53-1-1, Monthly Progress Report for January 1953, E. I. du Pont de Nemours and Company, 

February 6, 1953; SRDB Ref ID: 11419 

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DPSP 54-1-10, Monthly Progress Report for October 1954, E. I. du Pont de Nemours and Company, 

November 16, 1954; SRDB Ref ID: 11458 

DPSP 54-1-11, Monthly Progress Report for November 1954, E. I. du Pont de Nemours and 

Company, December 20, 1954; SRDB Ref ID: 11459 

DPSP 55-1-8-DV, Monthly Progress Report, August 1955, E. I. du Pont de Nemours and Company, 

September 19, 1955; SRDB Ref ID: 11473 

DPSP 55-1-9-DV, Monthly Progress Report, September 1955, E. I. du Pont de Nemours and 

Company, October 21, 1955, SRDB Ref ID: 11476 

DPSP-55-454-2, Savannah River Plant History, Separations Areas, July 1954 through December 

1972, E. I. du Pont de Nemours and Company, April 1973; SRDB Ref ID: 48560 

DPSP-55-454-3, Savannah River Plant History, Raw Materials Area, July 1954 through December 

1972, E. I. du Pont de Nemours and Company, February 1974; SRDB Ref ID: 48563 

DPSP 56-1-8, Monthly Progress Report, August 1956, E. I. du Pont de Nemours and Company, 

September 18, 1956; SRDB Ref ID: 12057 

DPSP 56-1-9, Monthly Progress Report, September 1956, E. I. du Pont de Nemours and Company, 

October 18, 1956; SRDB Ref ID: 16859 

DPSP 57-1-1, Monthly Progress Report, January 1957, E. I. du Pont de Nemours and Company, 

February 18, 1957; SRDB Ref ID: 11575 

DPSP 57-1-2, Monthly Progress Report, February 1957, E. I. du Pont de Nemours and Company, 

March 20, 1957; SRDB Ref ID: 11576 

DPSP 57-1-7, Monthly Progress Report, July 1957, E. I. du Pont de Nemours and Company, August 

19, 1957; SRDB Ref ID: 11601 

DPSP 57-1-8, Monthly Progress Report, August 1957, E. I. du Pont de Nemours and Company, 

September 19, 1957; SRDB Ref ID: 11602 

DPSP 57-1-9, Monthly Progress Report, September 1957, E. I. du Pont de Nemours and Company, 

October 18, 1957; SRDB Ref ID: 11603 

DPSP 57-1-11-DV, Monthly Progress Report, November 1957, E. I. du Pont de Nemours and 

Company, December 18, 1957; SRDB Ref ID: 11605 

DPSP 58-1-12-DV, Monthly Progress Report, December 1958, E. I. du Pont de Nemours and 

Company, January 25, 1959; SRDB Ref ID: 11615 

DPSP 59-1-6, Progress Report June 1959, E. I. du Pont de Nemours and Company, July 21, 1959; 

SRDB Ref ID: 11628 

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DPSP 60-1-10-DV, Progress Report, October 1960, E. I. du Pont de Nemours and Company, 

November 17, 1960; SRDB Ref ID: 11786 

DPSP-61-1-2, Progress Report – February 1961, H. L. Butler, E. I. du Pont de Nemours & Co., 

March 16, 1961; SRDB Ref ID: 12453 

DPSP 61-1-10-DV, Progress Report, October 1961, E. I. du Pont de Nemours and Company, 

November 20, 1961; SRDB Ref ID: 11796 

DPSP-67-1281, Polonium Production in Curium-II, P. L. Gray, June 30, 1967; SRDB Ref ID: 49819 

DPSPU-58-030-011, Rapid Determination of Mixed Beta-Gamma Radionuclides in Urine, A. L. Boni, 

E. I. du Pont de Nemours and Company; SRDB Ref ID: 49225 

DPSPU-58-030-014, Radiation Protection Considerations in a Heavy Water Moderated Reactor, 

Savannah River Plant; E. I. du Pont de Nemours & Co.; September 1958; SRDB Ref ID: 48566 

DPSPU-60-11-7, Sensitive Methods for Determination of Tritium in Water and Urine, F. E. Butler, 

June 1960; SRDB Ref ID: 49513 

DPSPU 62-30-5, Control of Tritium Health Hazards at the Savannah River Plant, W. C. Reinig, E. L. 

Albeneslus, Savannah River Plant, January 1962; SRDB Ref ID: 48558 

DPSPU-62-30-25A, Control of Tritium Hazards in Heavy-Water-Moderated Reactors, H. L. Butler, 

C. M. Patterson, September 1962; SRDB Ref ID: 49903 

DPSPU-63-30-037, Assessment of Tritium in Production Workers, F. E. Butler, Savannah River Plant, 

March 1964; SRDB Ref ID: 49905 

DPSPU 63-30-30, Determination of Actinides in Urine by Liquid Ion Exchange, F. E. Butler, E. I. du 

Pont de Nemours and Company, August 1963; SRDB Ref ID: 49584 

DPSPU 64-30-18, Separation of Actinides by Liquid Ion Exchange, Determination of U and Am-Cm 

in Urine, F. E. Butler, September 1964; SRDB Ref ID: 49523 

DPSPU-75-30-7, Radiation Exposure Records Management, H. P. Boitar, Du Pont, December 1975; 

SRDB Ref ID: 8811 

NSEC008-P006, Startup to December 14, 1954, author unknown, date unknown; SRDB Ref ID: 

48929 

OCAS-IG-001, External Dos Reconstruction Implementation Guideline, Rev. 3; National Institute for 

Occupational Safety and Health (NIOSH); Cincinnati, Ohio; November 21, 2007; SRDB Ref ID: 

38864 

OCAS-PER-030, Savannah River Site TBD Revisions,; Rev. 0; National Institute for Occupational 

Safety and Health (NIOSH); Cincinnati, Ohio; December 12, 2007; SRDB Ref ID: 38872 

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OCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 0, 

National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; September 23, 

2004; SRDB Ref ID: 32022 

OCAS-TIB-006, Interpretation of External Dosimetry Records at the Savannah River Site (SRS), Rev. 

2; National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; October 4, 2007; 

SRDB Ref ID: 35409 

OCAS-TIB-007, Neutron Exposures at the Savannah River Site, Rev. 01; National Institute for 

Occupational Safety and Health (NIOSH); Cincinnati, Ohio; October 15 2007; SRDB Ref ID: 35675 

ORAUT-OTIB-0001, Technical Information Bulletin: Maximum Internal Dose Estimates for 

Savannah River Site Claims, Rev 00; Oak Ridge Associated Universities; July 15, 2003; SRDB Ref 

ID: 19407 

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

Procedures, Rev. 03 PC-1; Oak Ridge Associated Universities; December 21, 2005; SRDB Ref ID: 

20220 

ORAUT-OTIB-0014, Technical Information Bulletin: Assignment of Environmental Internal Doses 

for Employees Not Exposed to Airborne Radionuclides in the Workplace, Rev 00; Oak Ridge 

Associated Universities; June 22, 2004; SRDB Ref ID: 19432 

ORAUT-OTIB-0017, Interpretation of Dosimetry Data for Assignment of Shallow Dose, Rev. 01; 

Oak Ridge Associated Universities; October 11, 2005; SRDB Ref ID: 19434 

ORAUT-OTIB-0018, Technical Information Bulletin: Internal Dose Overestimates for Facilities with 

Air Sampling Programs, Rev 01, Oak Ridge Associated Universities; August 9, 2005; SRDB 

Reference ID 19436 

ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for External Dose Assignment, Rev. 01; Oak 

Ridge Associated Universities; October 7, 2005; SRDB Ref ID: 19440 

ORAUT-OTIB-0032, External Coworker Dosimetry Data for the Savannah River Site, Rev. 00, PC-1; 

Oak Ridge Associated Universities; November 7, 2006; SRDB Ref ID: 29964 

ORAUT-OTIB-0052, Parameters to Consider When Processing Claims for Construction Trade 

Workers, Oak Ridge Associated Universities; Rev. 00 PC-1, January 16, 2007; SRDB Ref ID: 29978 

ORAUT-OTIB-0060, Technical Information Bulletin: Internal Dose Reconstruction, Rev 00, Oak 

Ridge Associated Universities; February 6, 2007; SRDB Ref ID: 29984 

ORAUT-PROC-0061, Occupational X-Ray Dose Reconstruction for DOE Sites, Rev. 01; Oak Ridge 

Associated Universities; July 21, 2006; SRDB Ref ID: 29987 

ORAUT-OTIB-0066, Calculation of Dose from Intakes of Special Tritium Compounds, Rev. 00; Oak 

Ridge Associated Universities; April 26, 2007; SRDB Ref ID: 31421 

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ORAUT-OTIB-0075, Use of Claimant Data Set for Coworker Modeling, Oak Ridge Associated 

Universities; No date yet; SRDB Ref ID: to be published in 2008 

ORAUT-TKBS-0003, Savannah River Site, Rev. 02; Oak Ridge Associated Universities; April 5, 

2005; SRDB Ref ID: 20176 

PNL-6301, 1987, Neutron Dose and Energy Spectra Measurements at Savannah River Plant, L. W. 

Brackenbush, et al; Pacific Northwest Laboratory; August 1987; SRDB Ref ID: 263 

PNNL-13421, A Compendium of Transfer Factors for Agricultural and Animal Products (June 2003), 

K. Rhoades, B. A. Napier, D. L. Strenge, L. H. Staven, PNNL, June 2003; SRDB Ref ID: 23593 

QH000-0385-92-002-P003, SRS Dosimetry Quarterly Report for Period Ending September 1970, 

Health Physics Department, October 21, 1970, SRDB Ref ID: 50075 

QH230-0521-93-004-P006, SRS Dosimetry Quarterly Report for Third Quarter 1965, October 8, 

1965; SRDB Ref ID: 50458 

WSRC-IM-90-139, Internal Dosimetry Technical Basis Manual, Rev 1; December 20, 1990; SRDB 

Ref ID: 11266 

WSRC-IM-94-39, History of 232-F, Tritium Extraction Processing (U), G. W. Blackburn, August 

1994; SRDB Ref ID: 12094 

WSRC-MS-2000-00061, Savannah River Site Fiftieth Anniversary, 50 Years of Excellence in Science 

and Engineering at the Savannah River Site, J. W. Joseph, C. J. Banick, May 17, 2000; SRDB Ref ID: 

45046 

WSRC-RP-95-234, A History of Personnel Radiation Dosimetry at the Savannah River Site (U), G. A. 

Taylor, K. W. Crase, T. R. LaBone, W. H. Wilkie, May 1995; SRDB Ref ID: 12098 

WSRC-SA-11, Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis 

Report, Executive Summary, Chapters 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, 16, 17, C. B. 

Shedrow, August 1997; SRDB Ref ID: 48970 

Arcano, 1994, Review of Facilities and Unirradiated Fuel Storage at the Reactor Materials Area of 

the Savannah River Site, March – April 1994, J. T. Arcano Jr., Defense Nuclear Facilities Safety 

Board, May 19, 1994; SRDB Ref ID: 44906 

Barg, 1996, Litco International Report: Technical Basis for the SMC Bioassay Program, Rev. 1, D. C. 

Barg, B. L. Anderson, January 16, 1996; SRDB Ref ID: 27055 

Bio-Assay Procedure, 1958, Bio-Assay Procedure for Tritium and Thorium, Ellett, J. D., February 13, 

1958; SRDB Ref ID: 48462 

Bio-Assay Samples, 1957, Bio-Assay Samples 100-K Area Shutdown, J. A. Harper, September 5, 

1957; SRDB Ref ID: 245 

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Bioassay Logbook, 1954-57, Bioassay Fission Products Controls Record Book, 12-20-1954 Thru 1-4-

1957; no attribution; 1954-1957; SRDB Ref ID: 48920 

Bebbington, 1990, History of Du Pont at the Savannah River Plant, W. P. Bebbington, 1990; SRDB 

Ref ID: 11249 

Branche, 2008, Letter to Petitioner from NIOSH Office of the Director concerning Administrative 

Review Panel Findings, C.M. Branche, August 14, 2008, OSA Ref ID: 106860 

Cardarelli, date unknown, Power Point Presentation: Study of Radiation Doses-Potential 

Consequences of Excluding Work-Required X-Ray Exposures When Computing Cumulative J. 

Cardarelli, date unknown; SRDB Ref ID: 11648 

Clemons, 1993, SRS ALARA Assessment June 28 – July 2, 1993, L. Clemons and R. Zavadoski, 

Defense Nuclear Safety Board, September 21, 1993; SRDB Ref ID: 48950 

DOE, 2008, Environmental Management Savannah River Site, U.S. Department of Energy, 

Environmental Management website; http://www.em.doe.gov/bemr/bemrsites/sars.aspx?182,128; 

9/19/2008; SRDB Ref ID: 48954 

Ellett, 1958, Bio-Assay Procedure for Tritium and Thorium, J. D. Ellett and C. M. Patterson; February 

13, 1958; SRDB Ref ID: 48462 

Exposure Evaluation, 1964, Exposure Evaluation and Dosimetry Program Review, H. L. Butler, 

January 23, 1964; SRDB Ref ID: 48904 

FAS, 1998, Heavy Water Production, Federation of American Scientists; October 21, 1998; SRDB 

Ref ID: 48957 

Fix, 1994, An Assessment of Bias and Uncertainty in Recorded Dose from External Sources of 

Radiation for Workers at the Hanford Site, J. J. Fix, E. S. Gilbert, W. V. Baumgartner; August 1994; 

SRDB Ref ID: 305 

Fix, 2006, Record of Phone Discussion with Ken Crase – Summary of Discussion of Selected Cohen 

& Associates SRS Site Profile Comments, J. Fix; August 15, 2006; SRDB Ref ID: 27056 

Hall, 1967, A Comparison of Lithium Fluoride and Film for Personnel Dosimetry, R. M. Hall, C. N. 

Wright; February 1, 1967; SRDB Ref ID: 11274 

Health Physics, 1954, Health Physics Area Survey Report April 1954, C.M. Patterson, May 05, 1954; 

SRDB Ref ID: 49710 

Health Physics, 1957, Health Physics Bio-Assay Program 100-L Extended Shutdown, J. L. Elrod, J. 

W. Harper, August 26, 1957; SRDB Ref ID:49759 

Health Physics Aspects, 1958, Health Physics Aspects of Reactor Operations 1956 versus 1957, H. A. 

McClearen, January 17, 1958; SRDB Ref ID: 49506 

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Huntoon, 1954, Test Authorization 3-138: AIS1 Canning Thorium Slugs, R. T. Huntoon, J. R. Driear; 

August 27, 1954; SRDB Ref ID: 49207 

Logbook P006, 1956, Thorium Record Book, 2-28-1956 Thru 11-7-1956, Author unknown; Savannah 

River Site; SRDB Ref ID: 49573 

Morris, 2008, SRS Thorium Processing Timeline to Support SRS SEC Evaluation Report, Robert 

Morris, September 26, 2008; SRDB Ref ID: Not yet in SRDB 

Personal Communication, 2008a, Personal Communication with current worker, Summary of 

Discussion of Selected Cohen & Associates SRS Site Profile Comments; Telephone Interview by 

ORAU Team; August 14, 2006; SRDB Ref ID: 27056 

Personal Communication, 2008b, Personal Communication with former worker, Use of NTA Neutron 

Personnel Dosimetry in 1955-1968 at SRS; Telephone Interview by ORAU Team; March 9, 2008; 

SRDB Ref ID: 48459 

Personal Communication, 2008c, Personal Communication with current worker, SRS Incident 

Database, Special Hazards Investigations Database, HPAREH Database and Intake Registry; 

Telephone Interview by ORAU Team; April 3, 2008; SRDB Ref ID: 45062 

Personal Communication, 2008d, Personal Communication with engineer concerning thorium, 

Telephone Interview by ORAU Team; August 12, 2008; SRDB Ref ID: Not yet released by DOE 

Personal Communication, 2008e, Personal Communication with metallurgical lab technician, 

Telephone Interview by ORAU Team; August 11, 2008; SRDB Ref ID: Not yet released by DOE 

Personal Communication, 2008f, Personal Communication with electrician, Telephone Interview 

ORAU Team; August 11, 2008; SRDB Ref ID: Not yet released by DOE 

Personal Communication, 2008g, Personal Communication with former Health Physics manager, 

Telephone Interview ORAU Team; August 11, 2008; SRDB Ref ID: Not yet released by DOE 

Peterson, 2002, Tritium Doses from Chronic Atmospheric Releases: A New Approach Proposed for 

Regulatory Compliance, Volume 82, S. R. Peterson and P. A. Davis; February 2002; SRDB Ref ID: 

46071 

Santos, 2007, Decommissioning of the 247-F Fuel Manufacturing Facility at the Savannah River Site, 

J. K. Santos and S. M. Chostner; SRDB Ref ID: 48948 

SEC, 2008a, SEC Worker Outreach Meeting for the Savannah River Site, National Institute for 

Occupational Safety and Health; May 22, 2008, 1 PM; SRDB Ref ID: Released by DOE; not yet in 

SRDB 

SEC, 2008b, SEC Worker Outreach Meeting for the Savannah River Site, National Institute for 

Occupational Safety and Health; May 22, 2008, 6 PM; SRDB Ref ID: Not yet released by DOE 

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48960 

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Singh, 2007, Excerpts and Notes from the Site Incident Databank at SRS, L. P. Singh, DOE, April 11, 

2007; SRDB Ref ID: 32167 

SRS, 1959, SRS Personnel Meter Audit Record 1959-1963, author unknown, March 19, 1959; SRDB 

Ref ID: 11291 

SRS, 2008, SRS Historical Highlights of the Savannah River Site, SRS, May 20, 2008; SRDB Ref ID: 

SRS at Fifty, 2002, Savannah River Site at Fifty – Reactor On, M. B. Reed, M. T. Swanson, S. 

Gaither, J. W. Joseph, W. R. Harvey; U.S. Department of Energy and the Savannah River Site 

Operations Office; 2002; SRDB Ref ID: 24750 

SRS Defense, 2007, SRS Facts – Defense Waste Processing Facility, December 2007, Washington 

Savannah River Company; SRDB Ref ID: 48961 

SRS Dosimetry, 1959, Dosimetry Special Hazards Incident Investigations 001-099, April 29, 1959; 

SRDB Ref ID: 45086 

SRS Dosimetry, 1962, Dosimetry Special Hazards Incident Investigations 100-199, Endorf, F. H., 

July 3, 1962; SRDB Ref ID: 45093 

SRS Dosimetry, 1974a, Dosimetry Special Hazards Incident Investigations 200-299, Endorf, F. H., 

April 24, 1974; SRDB Ref ID: 45094 

SRS Dosimetry, 1974b, Dosimetry Special Hazards Incident Investigations 300-399, Endorf, F. H., 

July 07, 1974; SRDB Ref ID: 45096 

SRS Dosimetry, 1989, Dosimetry Special Hazards Incident Investigations 400-494, December 23, 

1989; SRDB Ref ID: 45098 

SRS F, 2008, SRS Facts – F Canyon, Washington Savannah River Company, January 2008; SRDB 

Ref ID: 48962 

SRS FB, 2007, SRS Facts – FB Line, Washington Savannah River Company, December 2007; SRDB 

Ref ID: 48963 

SRS H, 2007, SRS Facts – H Area Completion, June 5, 2007; SRDB Ref ID: 48964 

SRS Highlights, 2000, SRS Historical Highlights, DOE Savannah River Operations Office, September 

22, 2000; SRDB Ref ID: 11246 

SRS Intake, 2004, SRS Intake Database, May 12, 2004; SRDB Ref ID: 45076 

SRS Saltstone, 2007, SRS Facts – Saltstone Facility, Washington Savannah River Company, 

December 2007; SRDB Ref ID: 48482 

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SRS Spent, 2008, SRS Facts – Spent Nuclear Fuel at SRS, Washington Savannah River Company, 

January 2008; SRDB Ref ID: 49483 

SRS Tritium, 2008, SRS Facts - SRS Tritium Facilities, Washington Savannah River Company, 

January 2008; SRDB Ref ID: 48969 

Summary of Tritium, 1954, Summary of Tritium Contamination May 12-July 19, 1954, C. M. 

Patterson, July 21, 1954; SRDB Ref ID: 49701 

Thierry, 2002, Study of a Selection of 10 Historical Types of Dosimeter: Variation of the Response to 

Hp (10) with Photon Energy and Geometry of Exposure, I. Thierry, et al; April 19, 2002; SRDB Ref 

ID: 11296 

Washington Savannah River Company, 2008, SRS Facts – H Canyon, Washington Savannah River 

Company, January, 2008; SRDB Ref ID: 48965 

Wright, 1958, An Evaluation of the Savannah River Plant Film Badge System, C. N. Wright; February 

6, 1958; SRDB Ref ID: 48461 

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SEC-00103 11-14-08 SRS
 

Attachment 1: Savannah River Site Events Chronology 

Table A-1: Savannah River Site Events Chronology 

Dates Areas Activities 

February 1, 1951 400-D; K, C, L, 

P, R Reactors; 

200-F and 200­

H; 300-M 

Construction begins on on-site railroads; the Heavy Water 

Rework Facility (400-D Area) and its supporting coal-fired 

steam/power plant; the K, C L, P, and R Reactors (100 

Areas); F and H Canyons (300 Areas); the Fuel and Target 

Fabrication Facility (300-M); and the CMX and TNX 

technical development facilities. 

June 1952 320-M Normal assay uranium fuel slug cylinders manufactured at 

off-site locations begin to arrive at SRS. 

July 1952 M Several production support activities begin in the Fuel Target 

Manufacturing and Raw Materials Area. 

August 8, 1952 400-D D2O separation test operations begin at the Heavy Water 

Rework Facility. 

September 1952 305-M Graphite Test Pile reactor achieves first criticality. 

October 1952 400-D Full-scale operation of the D-400 Heavy Water Rework 

Facility begins. 

December 1952 305-M Heavy-water-moderated SRS prototype reactor, the “Process 

Development Pile” (PDP) achieves full operations status. 

January 1953 320-M 320-M fuel fabrication facility begins full-scale operations. 

First batch of SRS-produced “Mark I” finished fuel slugs are 

successfully tested at the end of the month. 

May 1953 400-D All heavy water production processes are operating at full 

capacity. 

July 1953 320-M First batch of SRS-produced control rods are successfully 

tested. 

December 2, 1953 105 Areas AEC authorizes start-up of L Reactor using enriched uranium 

fuel, and P, K, C Reactors to be loaded “for maximum 

plutonium production, with excess reactivity used for tritium 

production.” 

December 28, 1953 105-R R Reactor achieves first criticality using Mark I natural 

uranium fuel. 

February 20, 1954 105-P P Reactor achieves first criticality. 

June 1954 105-R First irradiated fuel containing plutonium is withdrawn from 

R Reactor. 

August 1954 313-M and 

320-M 

R&D and small-scale production activities began to can 

thorium metal slugs for inclusion in SRS reactor fuel. 

August 11, 1954 105-L, L Reactors achieves first criticality. 

August 14, 1954 105-K K Reactor achieves first criticality. 

November 1954 221-F Radioisotope separation work begins in F Canyon, A-Line 

Facility, and 221-F B-Line Facility. 

November 1954 200-F First delivery of plutonium product to AEC. 

December 1954 200-F HLW and LLW Tanks 1-7 placed in service. 

December 1954 E 76 acres between F and H Areas designated as SRS solid 

waste burial ground. 

February 1955 313-M and 

320-M 

SRS begins inspection and acceptance testing of thorium 

metal slugs canned by offsite vendors. 

March 28, 1955 105-C C Reactor achieves first criticality. All five SRS production 

reactors are now in operation. 

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SEC-00103 11-14-08 SRS
 

Table A-1: Savannah River Site Events Chronology 

Dates Areas Activities 

June 30, 1955 F-Area First high-level waste tank is filled. 

July 1955 221-H Purex radioisotope separation work begins in H Canyon. 

October 1955 200-F First tritium recovered from irradiated lithium-aluminum 

alloy targets. 

November 1955 217-F First delivery of tritium gas product to AEC. 

December 1955 F- and H-Areas Waste tanks 2-F, 9-H, 10-H, and 11-H put into service storing 

waste from Purex process. 

June 1956 232-H Tritium Facility goes into operation. 

November 1956 F- and H-Areas Waste tanks 3-F, 8-F, 12-H, and 13-H are filled. Four 

additional waste tanks placed in service in H Area. 

March 1957 F F Area operations suspended to allow a major upgrade of 

equipment to achieve greater capacity, including construction 

of the JB Line on the F Canyon roof. 

November 1957 H 14-H Waste Tank placed into service. 

December 1957 D 411-D Building placed in standby and half of the DW Plant 

converted to rework tritiated moderator. 

1958 100 P and R Reactor (Par) Pond, needed to provide cooling water 

for the P and R Reactors, is completed. 

1958 K-100 SRS initiates production of small quantities of Pu-238 for use 

by NASA as heat sources. 

1958 D Building 413-D (D2O extraction) and Plant E placed on 

standby. 

1959 F Purex operations resume in F Canyon. 

1959 F Waste tanks 5-F and 18-F receive high- and low-level wastes, 

respectively. Tanks 17-F and 20-F are constructed. 

1960 H SRS initiates construction of the Receiving Basin for Offsite 

Fuels (RBOF) 

1960 F Evaporation of waste started in F Area. 

1960 H and F 15-H and 20-F waste tanks put into service. 

1960 300 Hot press bonding slug production begins. 

1961 F and H F and H Canyon facilities begin alternating work schedules. 

1961 F Neptunium reprocessing begins in Building 235-F. 

1961 H Off-site tritium containers begin to arrive at Building 234-H. 

1961 F Wastes are placed in Tanks 4-F, 17-F, 19-F, and 21-F. 

1961 H Tanks 21-H and 24-H are placed into service. 

1962 All SRS begins installing air filters on some process emission 

sources. 

1962 B “Hector” - Heavy Water Components Test Reactor (770-U) 

begins operations. 

1962 H Tank 11-H receives HM high-level waste. 

1963 H Tank 12-H receives HM high-level waste. 

1963 F Building 235-F switched from neptunium to Cm-244 

fabrication. 

June 1964 R R Reactor shuts down. 

1964 B Heavy Water Components Test Reactor is shut down. 

1964 H Receiving Basin for Off-Site Fuels (RBOF) receives first 

shipment of spent fuel assemblies. 

1964 F Waste Tank 6-F receives Purex high-level waste. 

1964 H Receiving Basin for Off-Site Fuels (RBOF) sends waste to 

Tank 23-H. 

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SEC-00103 11-14-08 SRS
 

Table A-1: Savannah River Site Events Chronology 

Dates Areas Activities 

1964 H "Interim 23" U-233 separation from thorium fuel campaign 

begins. 

1965 H RBOF ships first off-site fuel to H-Canyon for processing. 

1965 F and H Simultaneous operations resume in both canyons. 

September 1965 H "Thorex" separation process to extract U-233 and thorium 

from thorium fuels begins. 

1967 F and H Waste transfer line connecting H and F Areas completed. 

1968 L L Reactor is shut down for design upgrades. 

1968 C C Reactor is shut down for repairs to tank wall. 

1969 H Tritium Facility begins recycling reservoirs. 

November 1969 H Final "Thorex" campaign is completed. 

1970 F Construction begins on Multi-Purpose Processing Facility 

(MPPF). 

1971 D Standby equipment in Buildings 411-D, 413-D, and E Plant is 

dismantled and sold for scrap. 

1971 H Waste Tanks 29-H and 32-H begin receiving wastes. 

1972 H Tank 31-H begins receiving waste concentrate. 

1972 F Canyon MPPF Building completed and put on standby. 

1973 All SRS adopts use of thermoluminescent dosimetry. 

1973 F Tank 34-F begins receiving waste concentrate. 

1974 H H Area processes low-enriched off-site fuel. 

April 1974 F Tank 33-F placed in service. 

February 1975 F Fire and explosion in A-Line necessitates shutdown of F 

Canyon 

August 1975 F F Canyon operations restart. 

1977 F Pu-238 production operations begin in the SRS Plutonium 

Fuel Form Facility (PUFF) 

1978 F Operations begin at the 238PuO2 Experimental Facility (PEF). 

1979 M Process Development Pile closed down in Building 777-M 

1980 L L Reactor restart initiative is launched. 

1980 All SRS begins major program to develop and use metal hydrides 

in its tritium production facilities. 

1980 SRTC (700 

Area) 

Thermal Cycling Absorption Process (TCAP) is invented at 

the Savannah River Technology Center. 

1981 M 305-M graphite test pile is shut down. 

1981 H Application development begins on the Thermal Cycling 

Absorption Process (TCAP) to separate hydrogen isotopes. 

1981 M M-Area Settling Basin clean-up project begins. 

1982 412-D Last part of the Heavy Water Rework Facility is permanently 

closed. 

1982 F and H Work begins to automate canyon processes. Distributed 

Control System (DCS) is installed in F-Area warm canyon. 

1982 All Tracking Atmospheric Radioactive Contaminants (TRAC) 

vehicle built. 

1983 H Construction of the Defense Waste Processing Facility 

begins. 

1983 SRTC Experimental TCAP achieves 97% D2/H2 purity. 

1984 SRTC TCAP is selected for application in the new Replacement 

Tritium Facility (RTF). 

1985 HB Line Production of Pu-238 is initiated. 

1985 100-C C Reactor shuts down for repairs (never restarts.) 

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SEC-00103 11-14-08 SRS
 

Table A-1: Savannah River Site Events Chronology 

Dates Areas Activities 

1985 100-L L Reactor restarts. 

1985 H New HB-Line begins producing Pu-238 for NASA’s deep 

space exploration program. 

1985 M Groundwater remediation system is constructed in M-Area. 

1986 Z Construction of Saltstone begins. 

1986 H Construction of Replacement Tritium Facility using the 

TCAP process begins. 

January 1987 F and H Construction begins on the F and H Effluent Treatment 

Project (ETP). 

1987 100 Areas New administrative power level limitations are placed on 

reactors due to concerns about emergency cooling systems. 

1987 F F-Area A-Line Operations shut down. 

August 1988 100-K K Reactor shuts down for maintenance. 

August 1988 100-L L Reactor shuts down for maintenance and upgrades. 

August 1988 100-P P Reactor shuts down for maintenance. 

October 1988 F and H Effluent Treatment Facility (ETF) begins treatment of 

wastewater from F- and H-Area separations facilities. 

1989 SRTC TCAP Prototype achieves 99% D2/H2 purity. 

1989 F 221-F B-Line shuts down. 

1990 100-K Construction of the K Reactor Cooling Tower begins. 

1990 Z Saltstone begins operations. 

1991 E Mixed Waste Management Facility in the Burial Ground 

Complex is closed. 

1991 M M-Area Settling Basin closure is completed. 

December 1991 100-L L Reactor placed in temporary shuts down status (never 

restarted.) 

July 1992 100-K K Reactor restarts briefly and is connected to new cooling 

tower, and is then placed on cold stand-by status. 

1992 H Non-radioactive test operations begin at the Tritium 

Replacement Facility. 

1993 100-K K Reactor permanently placed in cold-standby condition. 

1993 H Non-radioactive test runs initiated at the Defense Waste 

Processing Facility. 

1993 H Construction begins on Consolidated Incineration Facility. 

1993 F F Area A-Line restarts. 

1993 H Radioactive operations initiated at the Replacement Tritium 

Facility. 

1994 H 233-H, Replacement Tritium Facility, begins full-scale 

operations, including unloading gases from reservoirs 

returned from DOD, separating and purifying the useful 

tritium and deuterium, mixing the gases to exact 

specifications, and reloading the reservoirs. 

1996 H Vitrification work begins at the Defense Waste Processing 

Facility. 

1996 100-K K Reactor is placed on permanent cold shutdown status. 

1996 200-F Nuclear material stabilization work begins with restart of F 

Canyon facilities. 

1997 F First SRS high-level waste tanks are closed, starting with F20 Tank. 

1997 F F Area A-Line shuts down permanently. 

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SEC-00103 11-14-08 SRS
 

Table A-1: Savannah River Site Events Chronology 

Dates Areas Activities 

2000 100-K K Reactor Building is converted to K Area Materials Storage 

Area. 

July 2000 H Groundbreaking for Tritium Extraction Facility. 

March 2002 F Operations end in F Canyon. 

2003 F, H Areas SRS completes the transfer of 30,000 gallons of Am/Cm 

solutions from F Canyon to H-Area Tank Farm. 

February 2005 F FB Line operations are completed and discontinued. 

2005 H Construction of the Tritium Extraction Facility (TEF) is 

completed. 

October 14, 2005 F Site preparation begins for the Mixed Oxide (MOX) Fuel 

Fabrication facility, which will provide for the permanent 

removal of Pu from potential weapons uses. 

November 2006 H Tritium Extraction Facility begins to extract tritium from rods 

irradiated in Tennessee Valley Authority (TVA) reactors. 

January 2007 H TEF completes initial extraction of tritium from Tritium 

Producing Burnable Absorber Rods (TPBARs). 

August 1, 2007 F Construction of the MOX facility begins. 

Sources: SRS Highlights, 2000; Bebbington, 1990; DOE/EA-0948; Santos, 2007; Clemons, 1993; DOE, 2008; SRS, 2008; 

SRS Defense, 2007; SRS F, 2008; SRS FB, 2007; SRS H, 2007; Washington Savannah River Company, 2008; SRS 

Saltstone, 2007; SRS Spent, 2008; SRS Tritium, 2008; WSRC-SA-11 

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