October 2, 2007 

Mr. David Staudt 

Center for Disease Control and Prevention 

Acquisition and Assistance Field Branch 

Post Office Box 18070 

626 Cochrans Mill Road – B-140 

Pittsburgh, PA 15236-0295 

Re: Contract No. 200-2004-03805, Task Order 1: Transmittal of Draft Addendum to SCA-TRTASK1-0003, Status Report on the Resolution of the Savannah River Site Issues Resolution 

Matrix 

Dear Mr. Staudt: 

SC&A is pleased to submit to NIOSH and the Advisory Board its draft addendum to SCA-TRTASK1-0003, Status Report on the Resolution of the Savannah River Site Issues Resolution 

Matrix, dated October 2, 2007. This report has been subjected to a review for Privacy Actrelated information and edited accordingly, and is now cleared for unrestricted distribution. 

This document has not, however, been reviewed by the Advisory Board for factual accuracy or 

applicability within the requirements of 42 CFR 82 

If you have any comments or questions regarding this report, please contact me at 732-530-0104. 

Project Manager 

cc: P. Ziemer, PhD, Board Chairperson A. Makhijani, PhD, SC&A 

Advisory Board Members H. Behling, PhD, MHP, SC&A 

L. Wade, PhD, NIOSH M. Thorne, SC&A 

L. Elliott, NIOSH H. Chmelynski, SC&A 

J. Neton, PhD, NIOSH J. Fitzgerald, Saliant 

S. Hinnefeld, NIOSH J. Lipsztein, SC&A 

L. Homoki-Titus, NIOSH K. Robertson-DeMers, CHP, Saliant 

A. Brand, NIOSH S. Ostrow, PhD, SC&A 

J. Broehm, NIOSH K. Behling, SC&A 

C. Ellison, NIOSH T. Bell, Saliant 

L. Shields, NIOSH Project File (ANIOS/001/03) 

D. Sundin, NIOSH 

Sincerely, 

John Mauro, PhD, CHP 

1608 SPRING HILL ROAD, SUITE 400 • VIENNA, VIRGINIA • 22182 • 703.893.6600 • FAX 703.821.8236 
 

Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

Status Report on the Resolution of the Savannah River Site Issues Resolution Matrix 

Contract No. 200-2004-03805 

Task Order No. 1 

Addendum to SCA-TR-TASK1-0003
 

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

McLean, Virginia 22182 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

October 2007 

Disclaimer 

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

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

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

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

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

information only and that premature interpretations regarding its conclusions are unwarranted. 
 

Effective Date: 

October 2 , 2007 

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S. Cohen & Associates: 

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program 

Document No. 

Addendum to 

SCA-TR-TASK1-0003 

Effective Date: 

Draft – October 2, 2007 

Revision No. 

0 (Draft) 

STATUS REPORT ON THE RESOLUTION OF 

THE SAVANNAH RIVER SITE ISSUES 

RESOLUTION MATRIX 

Page 2 of 146 

Task Manager: ___________________ Date: _________

 Joseph Fitzgerald 

Supersedes: 

N/A 

Project Manager: _________________ Date: _________ 

John Mauro 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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TABLE OF CONTENTS 

Acronyms and Abbreviations ..........................................................................................................7 

1.0 Executive Summary ...........................................................................................................10 

1.1 Overview................................................................................................................11 

1.2 Matrix Issue Status.................................................................................................12 

2.0 Scope and Introduction ......................................................................................................17 

2.1 Review Scope.........................................................................................................17 

2.2 Review Approach...................................................................................................19 

2.3 Report Organization ...............................................................................................20 

3.0 Assessment Criteria and Methods......................................................................................22 

3.1 Objectives ..............................................................................................................22 

3.1.1 Objective 1: Completeness of Data Sources .............................................22 

3.1.2 Objective 2: Technical Accuracy..............................................................22 

3.1.3 Objective 3: Adequacy of Data.................................................................23 

3.1.4 Objective 4: Consistency among Site Profiles..........................................23 

3.1.5 Objective 5: Regulatory Compliance........................................................23 

3.2 Organization of TBD .............................................................................................23 

3.3 Review of Revision 03 of TBD .............................................................................24 

4.0 Issue/Comment Resolution Status .....................................................................................26 

4.1 Comment 1: Incomplete Assessment and Guidance pertaining to Recycled 

Uranium and Transplutonium Radionuclides ........................................................26 

4.1.1 Issue Description........................................................................................26 

4.1.2 Applicability to Revision 03 ......................................................................35 

4.1.3 Summary of Overall NIOSH Response .....................................................35 

4.1.4 Work Group Actions..................................................................................36 

4.1.5 Closure Status ............................................................................................37 

4.2 Comment 2: Beta/Gamma Adjustment and Uncertainty Factors .........................37 

4.2.1 Issue Description........................................................................................38 

4.2.2 Applicability to Revision 03 ......................................................................41 

4.2.3 Summary of NIOSH Response ..................................................................41 

4.2.4 Work Group Actions..................................................................................43 

4.2.5 Closure Status ............................................................................................43 

4.3 Comment 3: Inadequate Neutron-to-Photon Ratios..............................................44 

4.3.1 Issue Description........................................................................................44 

4.3.2 Applicability to Revision 03 ......................................................................50 

4.3.3 Summary of Overall NIOSH Response .....................................................50 

4.3.4 Work Group Actions..................................................................................51 

4.3.5 Closure Status ............................................................................................51 

4.4 Comment 4: Adequacy of the F- and H-Area Tank Farm Characterization .........51 

4.4.1 Issue Description........................................................................................51 

4.4.2 Applicability to Revision 03 ......................................................................58 

4.4.3 Summary of Overall NIOSH Response .....................................................58 

4.4.4 Work Group Actions..................................................................................59 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

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4.4.5 Closure Status ............................................................................................59 

4.5 Comment 5: Lack of Comprehensive Evaluation of Early Monitoring 

Program..................................................................................................................59 

4.5.1 Issue Description........................................................................................60 

4.5.2 Applicability to Revision 03 ......................................................................64 

4.5.3 Summary of Overall NIOSH Response .....................................................64 

4.5.4 Work Group Actions..................................................................................64 

4.5.5 Closure Status ............................................................................................65 

4.6 Comment 6: High-Five Approach Inconsistent with the Methodologies 

Recommended In 42 CFR Part 82 for Calculation of Internal Dose .....................65 

4.6.1 Applicability to Revision 03 ......................................................................66 

4.6.2 Summary of Overall NIOSH Response .....................................................67 

4.6.3 Work Group Action ...................................................................................67 

4.6.4 Closure Status ............................................................................................67 

4.7 Comment 7: Limitations Associated with the Assignment of Occupational 

Environmental Dose...............................................................................................67 

4.7.1 Issue Description........................................................................................68 

4.7.2 Applicability to Revision 03 ......................................................................73 

4.7.3 Summary of Overall NIOSH Response .....................................................73 

4.7.4 Work Group Action ...................................................................................73 

4.7.5 Closure Status ............................................................................................73 

4.8 Comment 8: Inadequate Evaluation of Special Tritium Compounds ...................74 

4.8.1 Applicability to Revision 03 ......................................................................78 

4.8.2 Summary of Overall NIOSH Comments ...................................................78 

4.8.3 Work Group Action ...................................................................................79 

4.8.4 Closure Status ............................................................................................79 

4.9 Comment 9: High-five Approach is based on Incomplete Data ...........................80 

4.9.1 Issue Description........................................................................................80 

4.9.2 Applicability to Revision 03 ......................................................................86 

4.9.3 Summary of Overall NIOSH Response .....................................................87 

4.9.4 Work Group Action ...................................................................................87 

4.9.5 Closure Status ............................................................................................87 

4.10 Comment 10: Technical Issues Associated with the High-Five Approach ..........88 

4.10.1 Issue Description........................................................................................88 

4.10.2 Summary of Overall NIOSH Response .....................................................89 

4.10.3 Work Group Action ...................................................................................89 

4.10.4 Closure Status ............................................................................................89 

4.11 Comment 11: ICRP 30 with Subsequent Use of ICRP 68....................................89 

4.11.1 Issue Description........................................................................................90 

4.11.2 Applicability to Revision 03 ......................................................................91 

4.11.3 Summary of Overall NIOSH Response .....................................................91 

4.11.4 Work Group Actions..................................................................................91 

4.11.5 Closure Status ............................................................................................91 

4.12 Comment 12: Solubility, Oro-Nasal Breathing, and Ingestion Assumptions .......92 

4.12.1 Solubility Assumptions ..............................................................................92 

4.12.2 Applicability to Revision 03 ......................................................................94 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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4.12.3 Summary of Overall NIOSH Response .....................................................94 

4.12.4 Work Group Actions..................................................................................94 

4.12.5 Closure Status ............................................................................................94 

4.13 Comment 13: Incidents and High-Risk Jobs Not Adequately Addressed ............94 

4.13.1 Issue Description........................................................................................94 

4.13.2 Applicability to Revision 03 ......................................................................98 

4.13.3 Summary of Overall NIOSH Response .....................................................98 

4.13.4 Closure Status ............................................................................................99 

4.14 Comment 14: Availability of Additional Source Documents and Data .............100 

4.14.1 Issue Description......................................................................................100 

4.14.2 Applicability to Revision 03 ....................................................................105 

4.14.3 Summary of Overall NIOSH Response ...................................................105 

4.14.4 Work Group Action .................................................................................105 

4.14.5 Closure Status ..........................................................................................105 

4.15 Comment 15: Ambiguous Dose Reconstruction Direction ................................106 

4.15.1 Issue Description......................................................................................106 

4.15.2 Applicability to Revision 03 ....................................................................106 

4.15.3 Summary of Overall NIOSH Response ...................................................107 

4.15.4 Work Group Action .................................................................................107 

4.15.5 Closure Status ..........................................................................................107 

4.16 Comment 16: Special Exposure Circumstances for Subcontractors and 

Construction Workers ..........................................................................................107 

4.16.1 Applicability to Revision 03 ....................................................................107 

4.16.2 Summary of Overall NIOSH Response ...................................................107 

4.16.3 Work Group Action .................................................................................107 

4.16.4 Summary ..................................................................................................108 

5.0 Overall Adequacy of the SRS Site Profile, Revision 03, as a Basis for Dose 

Reconstruction .................................................................................................................109 

5.1 Site Profile Improvements ...................................................................................109 

5.2 Satisfying the Five Objectives .............................................................................109 

5.2.1 Objective 1: Completeness of Data Sources ...........................................110 

5.2.2 Objective 2: Technical Accuracy............................................................111 

5.2.3 Objective 3: Adequacy of Data...............................................................115 

5.2.4 Objective 4: Consistency among Site Profiles........................................116 

5.2.5 Objective 5: Regulatory Compliance......................................................116 

5.3 Usability of the Site Profile for Its Intended Purposes ........................................116 

5.3.1 Ambiguous Dose Reconstruction Direction ............................................116 

5.3.2 Inconsistencies and Editorial Errors in the Site Profiles..........................116 

6.0 References........................................................................................................................118 

Attachment 1: NIOSH Technical Documents Considered During the Review Process.............133 

Attachment 2: WSMS Databank Statistics .................................................................................135 

Attachment 3: Recommended Criteria for Retrieval of Incidents from WSMS.........................145 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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LIST OF TABLES
 

Table 4-1. SRS Reactor Major Products ................................................................................ 26 

Table 4-2. Estimated Bone Surface Doses from Recycled Uranium to Workers at the 

Paducah Gaseous Diffusion Plant......................................................................... 28 

Table 4-3. Neutron-to-Photon Ratio Values Used as Surrogate Data for 

NTA Film Dosimeters........................................................................................... 46 

Table 4-4. Uncertainties Contributing to the Derivation of Neutron Dose by the 

Neutron-to-Photon Ratio Method ......................................................................... 47 

Table 4-5. Annual Average Number of Entries into the F- and H-Area Tank Farm
 

Data Bank in Various Periods ............................................................................... 54 

Table 4-6. Statistical Data from the tank farms Fault Tree Data Bank by Category. ............ 56 

Table 4-7. Some Airborne Particulate Mass Concentrations ................................................. 72 

Table 4-8. Summary of Dust Loading Studies Cites by Yu et al. (1993) (g/m3

) ................... 72 

Table 4-9. Average Hypothetical Intake Quantities for the High-Five Approach ................. 82 

Table 4-10. In-Vivo or Bioassay Information of Relevance Not Included in the HighFive Data............................................................................................................... 83 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

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ACRONYMS AND ABBREVIATIONS 

ABRWH 

or Advisory Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

Bq Bequerel 

CATI Computer-Assisted Telephone Interview 

CDC Centers for Disease Control and Prevention 

CEDE Committed Effective Dose Equivalent 

CFR Code of Federal Regulations 

Ci Curie 

DAF Dose Adjustment Factor 

DCF Dose Conversion Factor 

DOE Department of Energy 

dpm Disintegrations per minute 

DPSOP DuPont Savannah River Plant Operating Procedure 

DPSP DuPont Savannah River Plant Procedure 

DuPont E.I. DuPont De Nemours and Company 

EEOICPA Energy Employees Occupational Illness Compensation Program Act 

EPI Environmental Protection Institute 

EU Enriched Uranium 

FMPC Feed Materials Production Center 

GI Gastrointestinal 

HEPA High Efficiency Particulate Air 

HERB Health Related Energy Research Branch 

HPAREH Health Protection Annual Radiation Exposure History Database 

HT Tritiated Gas 

HTO Tritiated Water 

HVIS Health Visitor Information System 

ICRP International Commission on Radiation Protection 

IDR Internal Dosimetry Registry 

IMBA Integrated Modules for Bioassay Analysis 

IREP Interactive RadioEpidemiologic Program 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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Effective Date: 

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IRF Intake Retention Fraction 

keV Kiloelectron volts 

MDA Minimum Detectable Activity 

MeV Megaelectron volts 

MT Metal Tritide 

NCRP National Council on Radiation Protection and Measurement 

NIOSH National Institute for Occupational Safety and Health 

NRC Nuclear Regulatory Commission 

NTA Neutron Track Emulsion 

OBT Organically Bound Tritium 

OCAS Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities 

ORAUT Oak Ridge Associated Universities Team 

OTIB Oak Ridge Associated Universities Team Technical Information Bulletin 

pCi Picocuries 

PER Program Evaluation Report 

PNNL Pacific Northwest National Laboratory 

POC Probability of Causation 

RAC Risk Assessment Corporation 

RBA Radiation Buffer Area 

RCA Radiological Control Area 

RDZ Radiation Danger Zone 

RU Recycled Uranium 

RWP Radiation Work Permit 

SC&A S. Cohen and Associates 

SEC Special Exposure Cohort 

SHI Special Hazards Investigation 

SMT Special Metal Tritides 

SRP Savannah River Plant 

SRS Savannah River Site 

STC Stable Tritium Compounds 

TBD Technical Basis Document 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

 

 

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TIB Technical Information Bulletin 

TLD Thermoluminescent Dosimeter 

TLND Thermoluminescent Neutron Dosimeter 

UNH Uranyl Hexahydrate 

WSMS Washington Safety Management Solutions 

WSRC Westinghouse Savannah River Company 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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1.0 EXECUTIVE SUMMARY 

The original review of the Savannah River Site (SRS) site profile was conducted by 

S. Cohen and Associates (SC&A) and submitted to the Advisory Board on Radiation and Worker 

Health (ABRWH or Advisory Board) as a draft report on March 21, 2005. That review was 

based on Revision 02 of the SRS site profile. During its June 15, 2006, meeting in Washington, 

DC, the Board assigned SC&A to review a subsequent, updated version of that site profile, 

Revision 03, which had been issued in April 2005. The Board also empanelled a Work Group to 

guide discussions between the National Institute for Occupational Safety and Health (NIOSH) 

and SC&A on outstanding issues raised in the review of Revision 02 of the site profile [also 

referred to as the Technical Basis Document (TBD)]. All SC&A review material has been 

initially submitted to and coordinated with this Work Group, as well as with NIOSH. 

This report constitutes an evaluation of the progress made in resolving issues identified in the 

review of Revision 02 of the SRS TBD. The review focused on the extent to which Revision 03 

addressed and resolved the issues identified from the review of Revision 02 and the degree to 

which closure had been achieved.1

 SC&A also evaluated site-specific Technical Information 

Bulletins (TIBs) and procedures as a part of this review. Attachment 1 provides a complete list 

of documents considered during the evaluation. SC&A’s team of health physicists and technical 

personnel conducted the review from August 2006 to April 2007. The review included a Work 

Group meeting, technical calls, and a visit to SRS to review incident database information. 

NIOSH was provided a final opportunity to review the current status of issues prior to issuance 

of this report; however, its comments were delayed and will be forthcoming in the context of the 

upcoming SRS Work Group discussions. Accordingly, this review’s representation of 

outstanding SRS site profile issues (and NIOSH’s initial response at the time) does not 

necessarily reflect NIOSH’s position regarding these matters as of the date of the publication of 

this report, and it awaits NIOSH’s response and further Work Group deliberation and action. 

SC&A’s evaluation process included a review of the revised TBD, focused interviews with site 

experts, and review of documents previously retrieved from or subsequently requested of SRS. 

SC&A evaluated the TBD for completeness, technical accuracy, adequacy of data, compliance 

with stated objectives, and consistency with other site profiles, as stipulated in the SC&A 

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

SC&A is aware of NIOSH’s ongoing efforts to develop and issue Revision 04-E (Scalsky 2006) 

of the SRS site profile and has acknowledged additional information intended for inclusion in 

that version. (However, issues related to those intended improvements continue to be cited as 

“open” pending issuance of that revised site profile.) Likewise, additional guidance documents 

were being issued that, while not reflected in Revision 03 (Scalsky 2005) of the SRS site profile, 

would serve to mitigate some of the gaps and issues raised in this report. 

1

 The specific comments identified, the corresponding NIOSH response, and the subsequent Work Group 

actions originated with an issue resolution matrix informal draft report, entitled “Summary of Task 1 Savannah 

River Site Technical Basis Document Finding Matrix—Vertical Issues,” circulated and used at the August 22, 2006, 

Work Group meeting. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

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1.1 OVERVIEW 

Under the auspices of the Board’s Work Group, NIOSH and SC&A have made progress 

addressing a number of significant issues arising from SC&A’s original review of the SRS site 

profile, Revision 02. The revised TBD and/or various supporting documents have satisfactorily 

addressed a number of key technical issues identified in SC&A’s 2005 review, including the 

following: 

•	 To address the lack of guidance pertaining to shallow dose, NIOSH provided subsequent 

guidance in ORAUT-OTIB-0017, Interpretation of Dosimetry Data for Assignment of 

Shallow Dose (Merwin 2005), that SC&A found could be applied to shallow dose and 

nonuniform exposure at SRS. 

•	 Several generic issues, including assumptions related to internal dosimetry (oro-nasal 

breathing, solubility, and ingestion), complexity of implementing guidelines, and dose 

estimation for subcontractors and construction workers, are being addressed with the 

Board and SC&A in other venues. 

As noted previously, a number of issues seem amenable to resolution through proposed changes 

to be reflected by NIOSH in Revision 04-E (Scalsky 2006) of the site profile, including the 

following: 

•	 An expanded discussion of thorium, U-233, and Pu-242 sources and operations, now 

lacking in Revision 03 (Scalsky 2005), is planned for Revision 04-E (Scalsky 2006). 

•	 Further research by NIOSH and Oak Ridge Associated Universities (ORAU) is ongoing 

regarding SRS source terms and worker exposures to neptunium and curium, and 

Revision 04-E (Scalsky 2006) will reflect the results of this research. 

•	 Additional information has been added to draft Revision 04-E (Scalsky 2006) to better 

characterize the exposure potential of exotic radionuclide sources associated with the 

high flux program. 

However, the resolution of other key issues requires documentation, evaluations, or technical 

closure via the Work Group: 

•	 The adequacy of the characterization provided for the F- and H-Area Tank Farm remains 

a concern as both NIOSH and SC&A, through the Board’s Work Group, strive to obtain 

further information on radionuclide source terms through the Fault Tree Data Bank and 

other SRS incident databases. 

•	 The completeness of dose information for early SRS workers remains in question based 

on SC&A’s evaluation of the Health Protection Annual Radiation Exposure History 

Database (HPAREH) records file that is being relied upon for dose estimation. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

 

 

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•	 Although NIOSH has proposed an approach for conservatively estimating dose to 

workers from special tritium compounds (STCs), it falls short of characterizing who was 

exposed, where they were exposed, and in what time periods, all necessary parameters for 

dose estimation. 

•	 Concerns regarding the validity of neutron-to-photon ratios at SRS remain unresolved, 

given how neutron-to-photon ratios are assigned in Revision 04-E (Scalsky 2006). 

It is anticipated that the issuance of Revision 04-E (Scalsky 2006) of the TBD will be the means 

to address these issues and achieve final closure on those that are currently open. 

1.2 MATRIX ISSUE STATUS 

Comment 1: Recycled uranium. The site profile does not contain adequate guidelines for 

resolving uncertainties related to recycled uranium (RU) in ways that give the benefit of the 

doubt to the claimants. For instance, the TBD does not consider internal dose contributions for 

plutonium, other transuranics, or fission products. This is mitigated somewhat by the planned 

inclusion of updates in Revision 04-E (Scalsky 2006) of the TBD that deal directly with 

exposure from these radionuclides and bridge some of the gaps found in previous versions of the 

TBD. Among other actions, the Work Group has requested a timeline for the Oak Ridge 

Associated Universities Team Technical Information Bulletin (OTIB) regarding RU (publication 

pending), information on the ability to link to the bioassay program and extrapolate doses back in 

time, and the intended approach for thorium, U-233, and Pu-242. SC&A recommends that this 

issue remain OPEN with outstanding Work Group actions. 

Comment 2: Beta/Gamma correction factors. The issues associated with correction factors 

and uncertainties have not been satisfactorily resolved by Revision 03 or by planned additions in 

Revision 04-E (Scalsky 2006). The beta/gamma dosimeter adjustment factors and uncertainties 

applied underestimate the true exposure measured by the dosimeter. Correction factors applied 

to dosimeter results account for on-phantom calibration and do not consider uncertainty from 

field exposure conditions. The standard deviation for film dosimeters prior to 1971 is too low. 

SC&A agrees that ORAUT-OTIB-0017 (Merwin 2005) provides suitable guidance for the 

assignment of shallow dose. NIOSH provided references on the adjustment factors used in the 

TBD, and SC&A compared the uncertainty factors from the TBD with relevant site-specific 

workbooks and guides. SC&A recommends that this issue remain OPEN pending Work 

Group review. 

Comment 3: Neutron/photon ratios. The geometric mean and standard deviation that describe 

the post-1971 neutron-to-photon ratio are neither technically defensible nor likely to be claimant 

favorable to a large number of claimants. The neutron doses recorded by the thermoluminescent 

neutron dosimeter (TLND) between 1971 and 1995, as well as the pre-1971 neutron doses 

(derived from neutron-to-photon ratios), suffer from a high degree of uncertainty. The use of the 

95th percentile value for the TLND neutron dose of records is recommended. (Revision 04-E 

(Scalsky 2006) provides for the 95th percentile neutron-to-photon dose ratio to be applied to the 

recorded dose for likely noncompensable cases.) The Work Group requested from NIOSH a 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

 

 

 

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detailed description of neutron-to-photon ratio methods used at SRS. SC&A recommends that 

the issue remain OPEN pending Work Group review. 

Comment 4: Adequacy of radionuclide characterization. The adequacy of the F- and HArea Tank Farm characterization in Revision 03 of the TBD is questionable for use as dose 

reconstruction guidance. This is particularly true for early periods of operation, where primary 

records involving key operations and incidents are lacking. Moreover, no references are 

provided for the tank farm discussion in the TBD, and there is no analysis indicating how the 

conclusions were reached. The proposed Revision 04-E (Scalsky 2006) included (at the time of 

this review) only minimal changes that do not resolve key issues. The Work Group requested 

that NIOSH provide an expanded review of the adequacy of its SRS radionuclide list by 

reviewing additional source term information (e.g., the Fault Tree Data Bank). This led to a 

series of requests to the Department of Energy (DOE) for access to the Fault Tree Data Bank, 

culminating in a February 2007 visit to review firsthand selected files of what turned out to be 

the Washington Safety Management Solutions (WSMS) data file of likely different pedigree. A 

series of actions, provided in Section 4.13 of this report, stemmed from this onsite review. 

SC&A recommends that this issue remain OPEN with additional actions identified. 

Comment 5: Use of early monitoring data. The adequacy of early worker monitoring data is 

questionable and requires further investigation. This issue will not be resolved by an inventory 

of records provided in claimant files. Additional validation of the HPAREH database as the 

exclusive source for external coworker dose determination is necessary to demonstrate that such 

early data are adequate for use. The Work Group requested that NIOSH provide a description of 

categories of workers and timeframes during which workers were not appropriately monitored 

during the early phases of beginning operations; SC&A was requested to provide an inventory of 

data sources used in the preparation of individual dosimetry data submitttals. SC&A 

recommends that this issue remain OPEN pending further Work Group review. 

Comment 6: Validity of high-five approach. Revision 03 continues to use the “high-five” 

approach [by reference to ORAUT-OTIB-0001 (Brackett 2003)] and, in SC&A’s view, would 

remain inconsistent with the methodologies recommended in Title 42, Part 82, Methods for 

Radiation Dose Reconstruction Under the Energy Employees Occupational Illness 

Compensation Program, of the Code of Federal Regulations (42 CFR Part 82), for the 

calculation of internal dose. NIOSH has indicated its intention to update the high-five approach 

and base the revised calculations on bioassay data, rather than data in the SRS Internal 

Dosimetry Registry (IDR); however, this update has not yet been released and referenced in 

Revision 04-E (Scalsky 2006). SC&A recommends that this issue remain OPEN pending 

further Work Group review. 

Comment 7: Onsite atmospheric dispersion and resuspension. The method used to 

reconstruct doses to unmonitored outdoor workers due to airborne emissions employs an 

atmospheric dispersion model, assumptions, and resuspension factors that do not appear to be 

claimant favorable and are not entirely appropriate for this class of problem. Revision 03 and 

proposed Revision 04-E (Scalsky 2006) provide no clarification on these issues. SC&A 

recommends that this issue remain OPEN pending further Work Group review. 

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Comment 8: Special tritium compounds. The TBD does not adequately address potential 

exposures of workers handling tritium and performing decontamination and decommissioning to 

STCs, including OBT and stable metal tritides. NIOSH did issue ORAUT-OTIB-0066, 

Revision 00, Calculation of Dose from Intakes of Special Tritium Compounds (LaBone 2006), 

but this guidance requires knowledge of site-specific processes. For SRS, NIOSH has provided a 

proposed site-specific dose estimation methodology for stable tritium compounds (STCs) at 

SRS; however, they have not verified the timeframe or location where STCs were handled or the 

types and quantities of STCs handled at SRS. An evaluation of the adequacy of the dose 

estimation methodology cannot be carried out without this key information, which the Work 

Group has requested. SC&A recommends that this issue remain OPEN pending further 

Work Group review. 

Comment 9: Completeness of intake values for the high-five approach. Intake values for the 

high-five approach are derived from an average of the highest five intakes identified by NIOSH. 

The values used to derive the average were obtained from a limited review of the SRS IDR, 

which does not contain all intakes occurring at SRS. The Work Group requested actions on the 

part of both NIOSH and SC&A, including completing an update of high-five intakes using new 

models, reviewing the Fault Tree Data Bank for applicability, and reviewing additional 

information available on 3x5 visitor cards. SC&A recommends that this issue remain OPEN 

with additional actions identified. 

Comment 10: Currency of International Commission on Radiation Protection (ICRP) 

methodology. The hypothetical intake, outlined in ORAUT-OTIB-0001 (Brackett 2003), uses 

the SRS intake quantities calculated using the methodology in ICRP 30, Limits for the Intake of 

Radionuclides by Workers (ICRP 1979). The average activity (nCi) is entered into the Integrated 

Modules for Bioassay Analysis (IMBA), and a dose is calculated based on the models in ICRP 

66, Human Respiratory Tract Model for Radiological Protection, and ICRP 68, Dose 

Coefficients for Intakes of Radionuclides by Workers. The use of surrogate data for internal dose 

for unmonitored workers adopted in the high-five approach for target organs that do not 

concentrate the radionuclides in question is not necessarily a maximizing approach for making 

dose estimates, contrary to the claim in the TBD. The Work Group requested a NIOSH update 

of high-five intakes using the new models. SC&A recommends that this issue remain OPEN 

with additional actions identified. 

Comment 11: Assignment of organ dose may not be sufficiently conservative. For internal 

dose calculations, the use of ICRP 30 methodology to calculate the intake with a subsequent use 

of ICRP 68 models to calculate the dose did not always result in the intended highest dose to an 

organ. Similarly, the appropriate solubility types between the two methodologies were not 

always paired consistently, resulting in discrepancies and a lack of claimant favorability. The 

dose reconstructor is directed to use surrogate radionuclides for radionuclides absent from the 

IMBA code. This issue has been partially resolved since the initial SC&A review with the 

inclusion of additional radionuclides in the latest version of the IMBA code, eliminating the need 

to use surrogate organs in many cases. Bioassay data rather than intake data should be used to 

calculate internal dose with the most current methodology available. As in Comment 10, 

resolution of this issue would be based on the results of a NIOSH update of its high-five intake 

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method using bioassay data and appropriate models. SC&A recommends that this issue 

remain OPEN with additional actions identified. 

Comment 12: Solubility, oro-nasal breathing, ingestion. Solubility, oro-nasal breathing, and 

ingestion should be carefully considered with regard to internal dose reconstruction. SC&A 

originally developed these points for the review in the Bethlehem Steel and Mallinckrodt 

Chemical Works site profile reviews, and they are applicable for all bioassay interpretations for 

the Energy Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA). 

The Work Group considered this a generic issue being addressed by NIOSH in a different venue. 

As a result, SC&A recommends that this issue be CLOSED; no further action necessary. 

Comment 13: Appropriate application of incident information. Incidents and high-risk jobs 

are not listed in the TBD or referenced to alert dose reconstructors to unique exposure 

conditions. Without a thorough reconciliation of the DOE exposure files against these separate 

incident data banks, NIOSH cannot be assured that all significant exposures from incidents are 

considered in relation to individual worker claims or to high-five estimates of maximal doses. 

For consistency among dose reconstructions, the reviewers concluded that the TBD should alert 

the dose reconstructor to special conditions when a deviation from the standard dose 

reconstruction methodology is needed. Work Group actions for resolving this issue included 

NIOSH requesting copies of the Special Hazards Investigations reports from DOE/Westinghouse 

Savannah River Company (WSRC), obtaining the “User’s Guide” for the WSMS database from 

DOE/WSRC, identifying other available incident databases, and researching the pedigree for the 

WSMS database reviewed during the February 2007 visit. SC&A recommends that this issue 

remain OPEN with additional actions identified. 

Comment 14: Completeness of HPAREH database. 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 coworker model. The integrity of the HPAREH file for use in coworker modeling is 

questionable given the absence of much of the data for workers terminating employment prior to 

1979. A basis for its appropriateness should be developed and discussed in either the TBD or the 

external coworker dose procedure. The Work Group requested that NIOSH review neutron 

logbooks referenced in the August 2006 meeting and evaluate the completeness of the HPAREH 

file used as a basis for the external dose coworker model. SC&A recommends that this issue 

remain OPEN with additional actions identified. 

Comment 15: Clarity of TBD information and guidance. Many of the sections of the TBD, 

especially Chapter 4 related to internal dosimetry, are very difficult to understand, and, together 

with the large array of TIBs and other Office of Compensation Analysis and Support/ORAU 

procedures, create a virtually impenetrable and complex array of guidelines. This situation lends 

itself to inconsistencies in the way in which dose reconstructions are performed and makes it 

difficult to verify the reliability and reproducibility of the dose reconstructions. NIOSH 

responded that a large number of TIBs and other guidelines have been developed to address the 

issues raised. SC&A recommends that this issued be CLOSED; no further action necessary. 

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Comment 16: Treatment of construction workers. The TBD does not currently include the 

special exposure circumstances for subcontractors and construction workers; however, NIOSH is 

aware of this issue. ORAUT-OTIB-0052, Parameters to Consider When Processing Claims for 

Construction Trade Workers (Chew et al. 2006) was developed to provide dose reconstruction 

guidance for trade workers. SC&A reviewed this procedure and submitted a draft report 

addressing this issue to NIOSH and the Board on July 30, 2007, as part of Task Order 3. SC&A 

recommends that this issue be CLOSED; no further action necessary. 

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2.0 SCOPE AND INTRODUCTION 

The original review of the Savannah River Site (SRS) site profile (SC&A 2005a) was conducted 

by S. Cohen and Associates (SC&A) and submitted to the Advisory Board on Radiation and 

Worker Health (ABRWH or Advisory Board) as a draft report on March 21, 2005. During its 

June 15, 2006, meeting in Washington, DC, the Board assigned SC&A to review a subsequent, 

updated version of that site profile, Revision 03, which had been issued in April 2005. The 

review was to focus on the extent to which Revision 03 addresses and resolves issues identified 

by SC&A from the review of Revision 02. This report presents the status of that review, which 

the SC&A team conducted from August 2006 to April 2007. 

The Board also empanelled a Work Group to guide the review, chaired by Michael Gibson, with 

Board members Mark Griffon, James Lockey, and Brad Clawson (the Work Group is now 

chaired by Mark Griffon). The Work Group arranged discussions with the National Institute for 

Occupational Safety and Health (NIOSH) and SC&A on selected outstanding issues, as well as 

information retrieval and onsite visits. All SC&A review material compiled under this review 

was submitted to and coordinated with this Work Group. The Work Group chairman briefed the 

full Board regularly on progress achieved in the SRS site profile issue resolution review. 

It is recognized that all site profiles are “works in progress” and are being revised. SC&A was 

aware of NIOSH’s ongoing efforts to develop and issue Revision 04-E (Scalsky 2006) of the 

SRS site profile and has acknowledged additional information intended for inclusion in that 

version. (However, issues related to those intended improvements continue to be cited as “open” 

pending issuance of that revised site profile.) Likewise, additional guidance documents were 

being issued that, while not yet reflected in Revision 03 of the SRS site profile, would serve to 

mitigate some of the gaps and issues raised in this report; where appropriate, these recent 

issuances have been so noted as well. NIOSH/Oak Ridge Associated Universities Team 

(ORAUT) proposed to the Work Group that Revision 04-E (Scalsky 2006) be used to close as 

many issues as possible and that comment resolution concentrate on issues that are still open. 

2.1 REVIEW SCOPE 

Under the Energy Employees Occupational Illness Compensation Program Act (EEOICPA) and 

federal regulations defined in Title 42, Part 82, Methods for Radiation Dose Reconstruction 

Under the Energy Employees Occupational Illness Compensation Program, of the Code of 

Federal Regulations (42 CFR Part 82), the Advisory Board is mandated to conduct an 

independent review of the methods and procedures used by NIOSH and its contractors for dose 

reconstruction. As a contractor to the Advisory Board, SC&A has been charged under Task 

Order 1 to support the Board in this effort by independently evaluating a select number of site 

profiles that correspond to specific facilities at which energy employees worked and were 

exposed to ionizing radiation. 

This report provides a review of Revision 03 of the SRS Technical Basis Document (TBD), 

ORAUT-TKBS-0003, Technical Basis Document for the Savannah River Site To Be Used for 

EEOICPA Dose Reconstructions, Revision 03, dated April 5, 2005 (Scalsky 2005), which 

superceded Revision 02 of ORAUT-TKBS-0003 issued in 2004, and supporting Technical 

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Information Bulletins (TIBs) and referenced technical information and databases. These 

included the following: 

Technical Basis Documents: 

•	 ORAUT-TKBS-0003, Technical Basis Document for the Savannah River Site To Be 

Used for EEOICPA Dose Reconstructions, Revision 03, April 5, 2005. (Scalsky 2005) 

Technical Support Documents: 

•	 OCAS-PER-001, Misinterpreted Dosimetry Records Resulting in an Underestimate of 

Missed Dose in SRS Dose Reconstruction, Revision 0, Office of Compensation Analysis 

and Support, Cincinnati, Ohio, September 8, 2003. (Neton 2003a) 

•	 OCAS-PER-002, Error in Surrogate Organ Assignment Resulting in an Underestimate of 

X-ray Dose in SRS Dose Reconstructions, Revision 0, Office of Compensation Analysis 

and Support, Cincinnati, Ohio, December 15, 2003. (Neton 2003b) 

•	 OCAS-PER-019, The Effect of Additional Neutron Dose Data from the Savannah River 

Site, Revision 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

May 18, 2007. (Allen 2007) 

•	 OCAS-TIB-006, Interpretation of External Dosimetry Records at the Savannah River 

Site (SRS), Revision 1, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

February 20, 2004. (Neton 2004) 

•	 OCAS-TIB-007, Neutron Exposures at the Savannah River Site, Revision 0, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, September 17, 2003. (Neton 

2003c) 

•	 ORAUT-OTIB-0001, Technical Information Bulletin: Maximum Internal Dose 

Estimates for Savannah River Site (SRS) Claims, Revision 0, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, July 15, 2003. (Brackett 2003) 

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

Revision 00 PC-1, Oak Ridge Associated Universities, Oak Ridge, Tennessee, 

November 7, 2006. (Merwin 2006) 

This list contains site-specific TIBs. Attachment 1 lists other generic TIBs that are applicable to 

SRS. 

Implementation guidance is also provided in “workbooks,” which have been developed by 

NIOSH for selected sites to provide more definitive direction to the dose reconstructors on how 

to interpret and apply TBDs, as well as other available information. The SRS-specific 

workbooks have been evaluated by SC&A under a separate task (Task 4) and are included in 

evaluations being submitted under that task. 

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SC&A, in support of the Advisory Board, has critically evaluated the revised SRS TBD to 

achieve the following: 

•	 Determine the completeness of the information gathered by NIOSH for the site profile 

with a view to assessing its adequacy and accuracy in supporting individual dose 

reconstructions. 

•	 Assess the technical merit of the data/information. 

•	 Assess NIOSH’s use of the data in dose reconstructions. 

SC&A’s review of the revised TBD, as with the prior review, focuses on the quality and 

completeness of the data that characterized the facility and its operations and the use of these 

data in dose reconstruction. SC&A conducted the review in accordance with Standard 

Operating Procedure for Performing Site Profile Reviews (SC&A 2004), which was approved 

by the Advisory Board. 

As a follow-up review, the evaluation is directed at reviewing previously identified issues and 

shortcomings identified by SC&A in Revision 02 of the TBD and determining whether these 

issues remain unresolved in Revision 03. As with other site profile reviews, this review is 

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

does not provide a rigorous quality control process whereby actual analyses and calculations are 

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

of the site profile that would be particularly influential in dose reconstructions, bridging 

uncertainties, or correcting technical inaccuracies. 

The TBD (for SRS, the term “TBD” is used interchangeably with “site profile”) serves as the 

primary site-specific guidance document used in support of dose reconstructions. These site 

profiles provide the health physicists who conduct dose reconstructions on behalf of NIOSH with 

consistent general information and specifications to support their individual dose reconstructions. 

SC&A prepared this report to provide the Advisory Board with an evaluation of whether and 

how the TBD and supporting documents can support dose reconstruction decisions. The criteria 

for evaluation include whether the TBD provides a basis for scientifically supportable dose 

reconstructions in a manner that is adequate, complete, efficient, and claimant favorable. 

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

workers were exposed and determine the level of exposure the worker received in that 

environment through time. The hierarchy of data used for developing dose reconstruction 

methodologies is first dosimeter readings and bioassay data, then coworker data and workplace 

monitoring data, and finally process description information or source term data. 

2.2 REVIEW APPROACH 

SC&A’s review of the revised TBD (Revision 03) and supporting documentation concentrated 

on determining the completeness of data collected by NIOSH, the adequacy of existing SRS 

personnel and environmental monitoring data, and the evaluation of key dose reconstruction 

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assumptions; all were performed in the context of the findings and issues identified in SC&A’s 

previous review of Revision 02 of the TBD. 

All review comments apply to Revision 03 of the SRS TBD, which is the most recent published 

version, although NIOSH has made available excerpts of the pending Revision 04-E (Scalsky 

2006) of the SRS TBD. SC&A is also aware of and is participating in ongoing informationgathering activities on site at SRS involving various incident exposure dose databases. In its 

2005 review of Revision 02 of the SRS site profile, SC&A found a number of shortcomings and 

needed clarifications that were identified in a series of findings and issue statements made in that 

draft report. SC&A highlighted these in an issue resolution matrix that was submitted to the 

Advisory Board shortly after the submission of the draft report. The matrix served to designate 

what issues (or parts of issues) were closed or open, and if open, what remained to be resolved. 

SC&A expanded and annotated this issue resolution matrix in detail on December 7, 2006, and 

July 3, 2007, as an issue status summary for comment and follow-up by the Work Group and 

NIOSH. 

On August 22, 2006, the Board’s Work Group held a meeting in Hebron, Kentucky, regarding 

the SRS review to identify issues and make assignments for follow-up. The Work Group 

addressed, in sequence, each finding identified by SC&A’s issue resolution matrix for SRS, with 

SC&A outlining the issue and its technical basis and NIOSH providing its technical position or 

current progress to resolve the issues. The Work Group, in some cases, identified followup 

actions for either NIOSH or SC&A to pursue to further resolution. 

This report is based on the July 3, 2007, issue status summary (derived from the original issue 

resolution matrix), updated to include actions stemming from the February 2007 visit as well as 

additional technical discussion regarding the open issues. NIOSH’s review of the SRS issue 

summaries compiled by SC&A was delayed, and comments will be forthcoming as a part of 

upcoming SRS Work Group discussions in October/November 2007. Accordingly, this review’s 

representation of outstanding SRS site profile issues does not necessarily reflect NIOSH’s 

position and awaits further Work Group deliberation and action. 

2.3 REPORT ORGANIZATION 

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

procedures prepared by SC&A and approved by the Advisory Board, this report is organized into 

the following sections: 

(1) Executive Summary 

(2) Scope and Introduction 

(3) Assessment Criteria and Methods 

(4) Issue/Comment Resolution Status 

(5) Overall Adequacy of the SRS Site Profile, Revision 03, as a Basis for Dose 

Reconstruction 

(6) References 

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This organization reflects the more proscribed nature of this follow-up review, and its focus on 

whether issues identified for Revision 02 of the TBD have been resolved in Revision 03 or has 

an acceptable intended treatment in draft Revision 04-E (Scalsky 2006). A number of the issues 

identified as “open” in this report are being addressed in Revision 04-E and appear to have 

satisfactory resolutions, but they cannot be closed until Revision 04-E is published. Those issues 

addressed by Revision 04-E are noted in this report where appropriate. 

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3.0 ASSESSMENT CRITERIA AND METHODS 

SC&A is charged with evaluating the approach set forth in the site profiles that is used in the 

individual dose reconstruction process. To do this, SC&A reviews the site profile documents for 

their completeness, technical accuracy, adequacy of data, consistency with other site profiles, 

and compliance with the stated objectives, as defined in SC&A Standard Operating Procedure 

for Performing Site Profile Reviews (SC&A 2004). SC&A identified 16 issues, listed in 

Attachment 3, in the original SC&A review from March 2005. The present review evaluates 

whether changes made in Revision 03 of ORAUT-TKBS-0003 (Scalsky 2005) resolved the 

issues identified in the review of Revision 02 of the SRS TBD. Items identified in this report 

may be applied to other facilities, especially facilities with similar source terms and exposure 

conditions. The review indicated that Revision 03 of the TBD does not resolve a majority of the 

16 concerns from the review of Revision 02. NIOSH/ORAUT has proposed a number of 

changes in Revision 04-E (Scalsky 2006) that will resolve some of the 16 issues identified once 

Revision 04-E is formally released. 

3.1 OBJECTIVES 

SC&A reviewed the site profile with respect to the degree to which it employs technically sound 

judgments or assumptions. SC&A compared each of the 16 findings from the March 2005 

review against Revision 03 of the TBD to determine whether the revision resolved the issue. In 

addition, the review identifies NIOSH assumptions that give the benefit of the doubt to the 

claimant. 

3.1.1 Objective 1: Completeness of Data Sources 

Objective 1 requires SC&A to identify principal sources of data and information that are 

applicable to the development of the site profile. The three elements examined under this 

objective include (1) determining if the site profile made use of available data considered 

relevant and significant to the dose reconstruction, (2) investigating whether other relevant/ 

significant sources are available but were not used in the development of the site profile, and 

(3) evaluating data compiled as a part of the comment resolution process. Additionally, SC&A 

evaluated records publicly available relating to SRS and records provided by site experts. 

3.1.2 Objective 2: Technical Accuracy 

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

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

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

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

occurred at SRS. The goal of this objective is to first analyze the data according to sound 

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

NIOSH/Oak Ridge Associated Universities Team (ORAUT) had analyzed available data, but 

SC&A found the technical approach used by NIOSH/ORAUT in the analysis of these data to be 

scientifically unsound or not necessarily claimant favorable, this would constitute a technical 

accuracy issue. 

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3.1.3 Objective 3: Adequacy of Data 

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

whether the data and guidance presented in the site profile are sufficiently detailed and complete 

to conduct dose reconstruction, and whether a defensible approach has been developed in the 

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

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

data that may influence the outcome of the dose reconstruction process. An example of an 

inadequacy in the data is the case of workers who appeared to have been exposed to neutrons but 

were not monitored for neutron exposures. 

3.1.4 Objective 4: Consistency among Site Profiles 

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

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

accomplish this objective, SC&A reviewed the previous analysis of consistency and documented 

the modifications. These modifications were compared to assumptions at other sites. 

3.1.5 Objective 5: Regulatory Compliance 

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

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

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

assurance policies and procedures used for the performance of dose reconstructions. 

3.2 ORGANIZATION OF TBD 

The Savannah River TBD, Revision 03, is divided into six sections, including an introduction, 

site process descriptions, and sections addressing internal dose, external dose, occupational 

medical occupational dose, and environmental occupational dose, as they pertain to historic 

occupational radiation exposure of SRS workers. 

Section 1.0, “Introduction,” explains the purpose and the scope of the site profile. SC&A was 

attentive to this section because it explains the role of the site profile in support of the dose 

reconstruction process. During the course of the review, SC&A was cognizant of the fact that 

neither EEOICPA nor 42 CFR Part 82, which implements the statute, requires the site profile. 

NIOSH developed site profiles as a resource for the dose reconstructors. Based on information 

provided by NIOSH personnel, SC&A understands that site profiles are living documents, which 

are revised, refined, and supplemented with TIBs as required, to help dose reconstructors. Site 

profiles are not intended to be prescriptive nor necessarily complete in terms of addressing every 

possible issue that may be relevant to a given dose reconstruction. Hence, the introduction helps 

in framing the scope of the site profile. As will be discussed later in this report, NIOSH may 

want to consider including additional qualifying information in the introduction to this and other 

site profiles describing the dose reconstruction issues that are not explicitly addressed by a given 

version of a site profile. 

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Section 1.0, as supplemented by Appendix A of the TBD, also includes process and activity 

descriptions of operations at SRS. This portion of the TBD is extremely important because it 

provides an overview of operations occurring on site through time and ultimately identifies 

sources of potential exposure. 

Section 2, “Occupational Medical Dose,” describes the methodology used reconstruct the 

medical exposures received by workers as a requirement for employment at SRS. SC&A 

reviewed this section for technical adequacy and consistency with other NIOSH procedures and 

other site profiles. 

Section 3, “Occupational Environmental Dose,” describes the methodology used to determine 

internal and external environmental dose at SRS. Environmental dose is assigned to 

unmonitored workers who were not likely to receive radiation exposure. Exposure of this type 

resulted from routine and episodic airborne emissions, resuspension or exposure to contaminated 

soil, and ingestion of contaminated food growing onsite. SC&A reviewed this section from the 

perspective of the source terms and atmospheric transport, deposition, and resuspension models 

used to derive the external and internal exposures to these workers. A section considering dose 

from the ingestion of contaminated foodstuffs was added to Revision 3 of the TBD and 

considered in the review. 

Section 4, “Occupational Internal Dose,” presents background information on in vitro and in 

vivo monitoring and air sampling at SRS. It describes methods of dose assignments from 

personnel monitoring data as well as the uncertainties associated with these data. An efficiency 

method referred to as the high-five approach is briefly described, as are the interference and 

uncertainties associated the internal monitoring. The detailed description of this high-five 

approach, which is a hypothetical intake, is available in ORAUT-OTIB-0001, Technical 

Information Bulletin: Maximum Internal Dose Estimates for Savannah River Site (SRS) Claims, 

dated July 15, 2003 (Brackett 2003). 

Section 5, “Occupational External Dose,” presents background information on beta, photon and 

neutron external monitoring at SRS. This section includes information on adjustment factors to 

be applied to SRS external dose values and uncertainties associated with these values. 

Section 6, “Trades workers,” is reserved in Revision 3 of the TBD. NIOSH has issued ORAUTOTIB-0052, Parameters to Consider When Processing Claims for Construction Trade Workers 

(Chew et al. 2006), for determination of dose to these workers. This procedure is under 

evaluation in a separate task. 

3.3 REVIEW OF REVISION 03 OF TBD 

The Advisory Board tasked SC&A with reviewing Revision 03 of the SRS TBD. This coincided 

with the comment resolution process. The Advisory Board formed a Work Group to guide 

discussions between NIOSH and SC&A on outstanding issues raised in the review of Revision 2 

of the TBD. SC&A prepared a matrix containing a summary of 16 issues identified in its review 

of Revision 02 and released it shortly after issuing the review. NIOSH had an opportunity to 

review this matrix and provide initial responses to the 16 issues. After the Work Group and 

SC&A received the formal response from NIOSH, the response was evaluated by SC&A and the 

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Work Group to determine whether the proposed actions resolved the issues. SC&A then 

evaluated Revision 03 of the TBD to determine whether the issue included in the matrix for 

Revision 02 still applied to Revision 03. For the benefit of the Work Group, SC&A prepared 

subsequent responses for discussion during the Work Group meetings and technical telephone 

calls. The Work Group, in cooperation with SC&A and NIOSH, either decided that an issue was 

closed or initiated follow-up actions for each issue. 

At the request of the Board’s Work Group, members of SC&A, NIOSH, ORAUT, and the 

Advisory Board conducted an onsite visit at SRS from February 28 to March 1, 2007. The 

purpose of the visit was to review, firsthand, what was thought to be the tank farms Fault Tree 

Data Bank cited in SC&A’s original findings regarding deficiencies in Revision 02 of the site 

profile as it pertained to the SRS tank farms. The three objectives of the onsite review were to 

determine the contents of the Washington Safety Management Solutions (WSMS) database, 

compare entries in this database to those from the tank farms Fault Tree Data Bank, and 

determine its usefulness in dose reconstruction. The database provided for review by SRS was 

actually the WSMS incident database, not the tank farms Fault Tree Data Bank. WSMS is a 

separate database developed by a former SRS employee for safety analysis of activities in the 

separations area. As a secondary task, SC&A conducted an interview in the presence of 

ORAUT, NIOSH, and Board members specifically dealing with special tritium compounds 

(STCs) at the site. 

Directed interviews were conducted by SC&A with a limited number of SRS workers. The 

information derived from these interviews was used to obtain additional supporting 

documentation or to substantiate comments described in the main text. A worker interview 

summary is not provided; however, interview information is integrated into the report and 

referenced. 

In July 2007, SC&A provided NIOSH, ORAUT, and the Work Group with an extended matrix 

including interchanges between NIOSH, ORAUT, SC&A, and the Work Group. At this time, 

NIOSH/ORAUT was offered an opportunity to provide additional comments prior to the 

issuance of this report. Although this report provides a status update, outstanding action items 

and issues that have not been satisfactorily resolved remain in the matrix. The process of 

comment resolution is ongoing until each issue has been resolved. Information from the 

matrices and conference calls has been provided in writing to members of the working group 

throughout the comment resolution process. Transcripts of Work Group meetings are published 

on the NIOSH Web site. 

Finally, it is important to note that SC&A’s review of Revision 03 of the TBD and its supporting 

TIBs focused on the areas identified in the matrix and is not exhaustive. 

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4.0 ISSUE/COMMENT RESOLUTION STATUS 

4.1	 COMMENT 1: INCOMPLETE ASSESSMENT AND GUIDANCE PERTAINING 

TO RECYCLED URANIUM AND TRANSPLUTONIUM RADIONUCLIDES 

The assessment and guidance pertaining to recycled uranium (RU) and some transplutonium 

radionuclides are incomplete. The revised site profile (Revision 03) does not contain guidelines 

for resolving uncertainties related to RU in ways that give the benefit of the doubt to the 

claimants. For example, the revised TBD does not consider internal dose contributions for 

plutonium, other transuranics, or fission products (e.g., Pu-239, Np-237, Tc-99, Ru-103, Rh-106, 

Sb-125, Zr-95, Nb-95, U-232, U-233, U-236, and U-237). Throughout the period when SRS and 

all other sites were producing and processing RU, few, if any, exposure limits were set for these 

radionuclides and no efforts were made to measure internal exposures to these isotopes among 

recycling workers. 

4.1.1	 Issue Description 

4.1.1.1 Recycled Uranium Impurities 

SRS processed significant quantities of RU, both from its own reprocessing plants and from 

other plants (DOE 1985; ERDA 1976; DuPont 1960; McCarty 2000; DOE 2001). For the SRS 

reactors, the major products are summarized by Boswell (2000) and reproduced below. 

Table 4-1. SRS Reactor Major Products 

Radioisotope When Produced Amount Application 

Plutonium-239 1954–1988 1000s kg Nuclear Weapons 

Tritium 1954–1988 100s kg Nuclear Weapons 

Uranium-233 1956–1968 100s kg Breeder Reactor Development 

Plutonium-238 1959–1988 100s kg Thermoelectric Generators 

for Space Exploration 

Plutonium-240 1958–1984 100s kg Target Material for 

Transplutonium Isotopes 

Plutonium-242 1964–1984 10s kg 

Cobalt-60 1956–1970 -

 66 mega curies Gamma Radiation Source 

Heat Generation Sources 

Thermal Electric Generators 

Curium-244 1962–1978 -

 12 kg Target for Production of 

Transplutonium Isotopes 

Polonium-210 1966–1969 -

 600 g Intense Radiation Source 

Californium-252 1965–1970 21 g Cancer Treatment, 

Oil-Well Logging, etc. 

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It is estimated that from 1959 to 1999, some 31,355 metric tons of uranium were shipped from 

SRS to other U.S. Department of Energy (DOE) sites, including (but not limited to) the gaseous 

diffusion plants in Oak Ridge, Tennessee, Paducah, Kentucky, and Portsmouth, Ohio; the Oak 

Ridge Y-12 Plant in Oak Ridge, Tennessee; and the Feed Materials Production Center (FMPC) 

in Fernald, Ohio (McCarty 2000). During this same time period, it is estimated that SRS 

received 54,544 metric tons of uranium from other sites, such as FMPC, DOE’s gaseous 

diffusion plants, and the Y-12 Plant (McCarty 2000). 

From 1961 to 1999, SRS processed approximately one-third of an estimated total of 250,000 

metric tons of RU in the DOE complex. SRS processed uranium metals, oxides, and solutions of 

various assays, including depleted uranium, natural uranium, and low-enriched and highly 

enriched uranium. Enriched uranium was also extracted from domestic and foreign research 

reactor spent fuel. In addition, from 1964 to 1969, thorium was recycled to produce U-233 

(McCarty 2000). During the peak period of the Cold War, SRS generated 2,000 to 3,000 drums 

of RU trioxide per year. During this same period of RU production and processing, 

approximately 300 workers handled these materials annually at SRS (McCarty 2000). 

The Revision 03 site profile review discussion of radionuclides found in RU is limited to the 

glossary and includes only uranium isotopes (Scalsky 2005, p. 136). Guidelines are not provided 

for resolving uncertainties related to RU in ways that give the benefit of the doubt to the 

claimants. For instance, the TBD does not consider internal dose contributions from plutonium 

or other transuranics, or fission products for uranium area workers. Recycled uranium is 

recovered from reprocessing plants after it has already been irradiated in a reactor one or more 

times. This creates uranium with radioisotopes that are not found in unirradiated uranium. 

Virgin uranium contains U-234, U-235, and U-238. Recycled uranium contains all three of 

these, as well as other isotopes of uranium, notably U-236, and traces of certain fission products 

and transuranic radionuclides. While the possible list of impurity radionuclides in RU is long, 

the main radionuclides potentally include Tc-99, Pu-238, Pu-239, Pu-240, Np-237, U-232, U233, and U-236 (DOE 1985). 

Throughout the period when SRS and all other sites were producing and processing RU, limited 

or no efforts were made to measure internal exposures from the impurities in RU. A preliminary 

analysis of the production, flow, and disposition of RU at SRS states the following (McCarty 

2000): 

SRS workers were not routinely monitored for exposure to plutonium, neptunium, 

or technetium that might have been present in the recycle uranium streams. 

To further compound the problem, DOE/OR-859, The Report of the Joint Task Force on 

Uranium Recycle Materials Processing (DOE 1985), states the following: 

A formal, technically sound, understood and accepted specification for maximum 

transuranic and fission product contaminants in uranium recycle material has 

probably never existed either within or between sites. 

Table A-2 of the SRS TBD provides a partial listing of radionuclides of concern for the 221-F 

Area A-Line facility, which converted depleted uranyl nitrate solution to uranium trioxide for 

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recycling (Scalsky 2005). However, this table does not equate with radionuclides of concern 

recommended by a special task force on RU convened in 1985 by DOE; these radionuclides 

include Pu-239, Np-237, Tc-99, Ru-103, Rh-106, Sb-125, Z-95r, Nb-95, U-232, U-233, U-236, 

and U-237 (DOE 1985). In fact, the presence of transuranic trace contamination was a large part 

of the reason that the uranium enrichment plants at Oak Ridge, Portsmouth, and Paducah were 

granted Special Exposure Cohort (SEC) status in EEOICPA. Such trace contamination has been 

shown (in the following excerpt) to have the potential for significant radiation doses, if the 

concentrations are high enough (DOE 2000, p. 77). 

Table 4-2. Estimated Bone Surface Doses from Recycled Uranium to Workers at the 

Paducah Gaseous Diffusion Plant 

(Committed Effective Dose Equivalent – CEDE) 

Average Air Concentrations Maximum Air Concentrations 

48.06 – 188 rems 599.24 – 2,238 rems 

SRS defines radionuclides of concern for the air monitoring and the bioassay program as follows 

(WSRC 2001): 

Although there may be many radionuclides present in a facility, typically only a 

few have the potential for delivering significant doses and they are usually quite 

obvious: uranium in uranium facilities, plutonium in plutonium facilities, and 

tritium and tritium facilities for example. Also, some radionuclides are important 

because they are relatively easily detected and can be used as tracers for the 

radionuclides that deliver the dose. Americium-241 in a plutonium facility is a 

good example. Radionuclides that deliver most of the dose and their tracers are 

referred to as radionuclides of concern. Air monitoring and bioassay programs 

are designated to detect these radionuclides. 

Radionuclides of concern are determined in the following manner: All 

radionuclides in a work area to which workers could be exposed are identified 

from waste certification records, contamination surveys, safety analysis reports, 

technical reports, the open literature, personal interviews, etc. The radionuclides 

in the area that deliver a cumulative dose fraction of more than 90% are deemed 

to be the radionuclides of concern and are considered for inclusion on the RWP. 

All other radionuclides may be ignored unless they are suitable for use as a 

tracer.... 

Radionuclides in a mixture resulting in less than 10% of the total dose are not considered 

significant in terms of air sampling and bioassay monitoring unless these radionuclides serve as a 

tracer for significant dose-producing radionuclides. In the case of an operational dosimetry 

program, this is justifiable as long as the site meets the intent of the regulations. In terms of a 

compensation program, the additional dose must be accounted for as different radionuclides 

concentrate in different organs of the body. Prior to excluding a radionuclide from analysis, it 

should be investigated in the context of all potential organs of interest. 

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Crase and LaBone (2000) evaluated the dose fractions from impurities, such as plutonium in RU, 

for materials processed and handled in SRS facilities. The source term was derived from a 

detailed assessment of the radionuclide mix in the 221-H waste stream (Elliott 1997). The 

relative activities of the radionuclides were normalized to an activity fraction and dose 

conversion factors were applied to the activity fractions to determine the committed effective 

dose equivalent (CEDE). The analysis used maximizing internal dose assumptions. Crase and 

LaBone (2000) summarized dose contributions from impurities in RU as follows: 

Dose fractions calculated from the radioisotope mix for the SRS uranium 

recovery facilities indicate that impurities do not contribute a significant fraction 

of the total dose. For the enriched uranium recovery facility, the total dose 

fraction due to impurities was less than 8%, assuming intake parameters that 

would maximize the internal dose contribution from impurities. For intake 

parameters that would maximize the internal dose from all radionuclides 

(including uranium), the impurity dose contribution is much less than 1%. In the 

depleted uranium recovery facility, impurities could contribute up to a maximum 

of 16% of the total dose, again assuming intake parameters that would maximize 

the internal dose from impurities. For intake parameters that would maximize the 

internal dose from all impurities (including uranium), the dose contribution from 

all impurities is much less than 1%. In none of the cases did any single 

radioisotope contribute as much as 10% of the total dose. Even using these 

conservation assumptions, the results support the SRS internal dosimetry practice 

of not monitoring SRS uranium workers routinely for plutonium and other 

actinides. 

The site clearly recognized the presence of impurities in RU. Crase and LaBone (2000) indicate 

that their analysis may not have been applicable to RU that may have been shipped to other 

nuclear facilities for additional processing or mixing. Based on the analysis completed by Crase 

and LaBone (2000), McCarty (2000) concluded the following: 

No evidence was found during the course of this study, which would indicate SRS 

recycled uranium presented any unusual challenge to radiation protection 

measures historically used at the site. 

This assertion does not inspire confidence that individual doses from trace contaminants in RU 

may not have been considerably higher. Data on fission product and transuranic impurities 

handled by workers are sparse at best (McCarty 2000): 

No authenticated copies of procedures from the majority of the processing period 

[involving the processing of recycled uranium] exist outside of the Records 

Management system, if they exist there. 

Reconstruction of doses to workers processing RU is made even more difficult because most of 

the laboratory personnel who performed analytical work on RU prior to the 1970s have long 

since retired. Thus, knowledge of changes in technology and analytical techniques, particularly 

during the 1950s and 1960s, is scant. 

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A preliminary review of historical records indicates that estimates of contaminant concentrations 

by Crase and LaBone (2000) of transuranic and fission product contaminants in RU could be 

tenuous. A DOE task force on RU reported in 1985 that the following occurred since the 

inception of the recycling program (DOE 1985): 

•	 There never existed “formal specifications on maximum permissible contaminant levels 

between reprocessing, intermediate and customer sites.” Rather, “...informal 

specifications in the form of ‘gentlemen’s agreements’ did evolve and have been in use 

since.” 

•	 Trace contaminant levels were increased without the proper review and concurrence that 

would have been required under formalized specifications. For example, in 1976, the 

maximum alpha activity specification from all transuranic elements of 

1,500 disintegrations per minute (dpm) per gram adopted by SRS in 1960 of total 

uranium was informally raised to 3,000 dpm/g uranium for shipment to the Fernald 

facility because of “the difficulty being experienced at SRP in attaining the 1,500 dpm g 

U specification.” 

•	 “Early SRP (1964–1972) returns [toY-12] based on 144 samples,” found that 10 samples 

exceeded the “gentlemen’s agreement,” with the highest at 180%. “Sample results over 

the most recent eight-year period (spanning 214 samples) indicate that 22 samples 

exceeded the informal specifications (the highest was 165%). It should be noted that 

SRP does not analyze for beta activity or recognize a beta specification.” 

The omission of transuranic and fission product isotopes from consideration in analyzing dose 

records of workers who handled RU may be a significant gap in internal dose for uranium 

facility workers, notably in those areas that were considered to have a “high potential” for 

worker contact with RU, including the following: 

•	 The FA-Line facility (in the 200 Area) in which uranium from the radiochemical 

separations operations was converted to trioxide. Workers involved in facility cleanup 

and removal of uranium trioxide (UO3) from the denitrator may have had the greatest 

contact with respirable RU particles. 

•	 Building 321-M, where casting and machining of RU was performed. In addition, 

building exhaust high-efficiency particulate air (HEPA) filter change-out activities may 

have also created a high potential for high airborne concentrations of RU. 

In summary, Revision 03 does not take into consideration transuranic and fission product 

contaminants in RU. Proposed Revision 04-E (Scalsky 2006) includes a discussion of RU and 

estimated impurities. NIOSH also indicated that a TIB concerning RU in the complex is in draft, 

but the document was not available for review. 

Dose reconstructors are instructed to include exposures from uranium impurities based on 

activity fractions in Table 4.6, ORAU-TKBS-0003, Revision 04-E (Scalsky 2006), for 

individuals having no impurity-specific bioassay (e.g., plutonium, neptunium). SC&A notes that 

Table 4.6 of ORAU-TKBS-0003, Revision 04-E (Scalsky 2006), contains estimates for 

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radionuclide impurities in RU based on estimates in waste streams (Crase and LaBone 2000). 

This table is derived from concentrations measured in waste streams and appears to be at odds 

with significantly higher contaminant levels set as product specifications at SRS for recycled 

uranyl hexahydrate (UNH) metal and low-enriched uranium, which were reported by the DOE 

1985 Joint Task Force on Recycled Uranium (DOE 1985, pp. 75–76). 

The Task Force found that several shipments of RU from SRS exceeded contaminant 

specifications by as much as 180 percent. (DOE 1985, pp. 47–48). The draft narrative in 

Revision 04-E (Scalsky 2006) does not address the processing of recycled UNH. It may be more 

claimant favorable and realistic to use impurity concentrations from shipments from SRS that 

exceeded product specifications to receiving sites. In view of this significant uncertainty, a 

further investigation of RU source term data should be completed to determine upper bounds of 

impurity concentrations and resulting doses. Other assays, such as metallurgical analyses, may 

assist in determining concentrations and relative uncertainties in these values. 

4.1.1.2 Pu-242 Programs 

While the primary products produced at SRS were plutonium and tritium, a variety of other 

isotopes were produced during the transplutonium program and for nonmilitary commercial uses. 

The transplutonium program started in the late 1950s and included the production and processing 

of Cf-252, Pu-242, Cm-244, and Am-243. The Curium I campaign produced Pu-242 from Pu239 using plutonium-aluminum assemblies. During the Curium II campaign, the material from 

Curium I was separated and purified. The plutonium was then refabricated into fuel and 

irradiated further to ultimately produce Cm-244. Because a high neutron flux was required to 

produce transplutonium isotopes, the site established the High Neutron Flux program in support 

of the curium programs. Furthermore, the High Neutron Flux program also resulted in the 

production of high specific-activity Co-60. Other activities at SRS included the thorium 

campaigns for the production of U-233 and the heat source programs that involved Pu-238, Po210, and Co-60. Cobalt-60 was later found to have uses in medicine and for sterilization. 

Special programs involved the production of other isotopes (e.g., Tm-170, Ir-192, Eu-152 and 

various isotopes of lanthanum) (Reed et al. 2002). Some of these radionuclides are considered in 

the dose reconstruction process, while others are not. 

In Revision 03, many of these isotopes are mentioned only as trace radionuclides or as a part of a 

routine mixture of product and/or waste. For example, Pu-242 is only mentioned as a trace 

contaminant (Scalsky 2005, p. 66) and not a material produced in its own right. The use of Pu242 as a radiobioassay tracer beginning in 1981 (Scalsky 2005, p. 66) may further complicate the 

detection of uptakes of plutonium. If a part of the recovered tracer in some cases was actually 

Pu-242 present in the bioassay sample, then the reported results would tend to underestimate the 

other plutonium isotopes present in addition to masking any intake of Pu-242. The TBD has not 

analyzed these campaigns to determine their potential influence on internal and external dose, the 

adequacy of the monitoring program with respect to these radionuclides, and the effect of these 

campaigns on isotopic ratios. This may be an important gap in the TBD. This gap affects 

bioassay as well as in vivo count interpretation for some groups of workers. The production of 

Pu-242 may also affect neutron dose calculations for Pu-242 production workers, as well as 

those in the target fabrication operations. 

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A brief description of the high weight percent (wt.%) Pu-242 source term was apparently added 

in the plutonium bioassay section of proposed Revision 04-E (Scalsky 2006). The effect of the 

higher wt.% Pu 242 was investigated with respect to its impact on plutonium bioassay yield 

calculations; however, the high Pu-242 campaigns ended in 1967, long before alpha 

spectrometry was used for plutonium bioassay. The impact on yield would have been less than 

10%, which is adequately accounted for in the overall geometric standard deviation of 3 used for 

intake uncertainty. 

In summary, Revision 03 does not consider potential contributions from exposure to high Pu-242 

containing plutonium. Revision 3 gives no justification for the absence of such a discussion. 

Proposed Revision 04-E (Scalsky 2006) contains a discussion of the production of Pu-242 during 

curium campaigns and provides information on the activity composition of a high Pu-242 

mixture. The default assumption for plutonium remains at 10-year-old 12% plutonium. SC&A 

will review proposed Revision 04-E (Scalsky 2006) of the TBD when it is released, if so directed 

by the Board. 

4.1.1.3 Thorium/U-233 

As a part of the breeder reactor program, U-233 was produced by the irradiation of thorium 

slugs. By 1956, U-233 was produced in small quantities. The production of U-232 impurities 

made fuel fabrication difficult because of the high gamma dose rates. This separation of fuel and 

target reduced the formation of U-232 impurities in 1965 to 3–6 parts per million. Several 

hundred kilograms of U-233 were produced through 1968. The Thorex process was used to 

recover U-233 and thorium from irradiation thorium targets at the F- and H-Canyon (WSRC 

2000). 

Revision 03 of the TBD provides a brief description of the U-233/thorium program in connection 

with the H-Canyon facility. Thorium was processed to recover the U-233. Attachment A 

identifies thorium as a radionuclide of concern at the 238PuO2 Fuel Form Facility and the 238PuO2 

Experimental Facility. U-233 is not mentioned as a radionuclide of concern for any facility. In 

fact, the TBD acknowledges that the SRS Internal Dosimetry Technical Basis Manual lists Th228 and Th-232 as radionuclides of concern (WSRC 2001). Information on thorium monitoring 

is limited. The TBD does not include guidance on the assignment of internal dose from thorium. 

Substantial revisions are proposed in Revision 04-E (Scalsky 2006) of the TBD. Changes 

include a description of the fabrication of thorium slugs, a discussion on exposure to U-233 

including impurities, and an added discussion of exposure to thorium including default intake 

assumptions. SC&A believes that this additional information will be helpful but provides the 

following cautionary notes. 

With respect to thorium, it appears to SC&A that the 0.8 curie (Ci) of Th-228 per 1 Ci of Th-232 

cited in this in Revision 04-E (Scalsky 2006) material may not be correct. It should be one-toone because Th-228 is not separated from Th-232 by any of the processing. For Ra-228, the 

ratio of 0.3 to 1 may be reasonable, but NIOSH needs to provide a better justification for this 

(i.e., why not 0.5? or 0.2?). NIOSH should indicate whether it has any data on the delay 

between storage and reprocessing or recovery from reprocessing and fuel fabrication. 

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Thorium urinalysis was seldom performed in the 1950s and 1960s (Scalsky 2006, p. 74), and the 

method used was not reliable. In fact, urinalysis was not a good monitoring tool until recently, 

when inductively coupled mass spectrometry started to be used, with very low limits of 

detection. The SRS site profile TBD (Scalsky 2006) states that in vivo monitoring was the main 

monitoring method. It also mentions that, in 1990, thorium was measured through the 106 L-xray (abundance 0.12%). This is only possible using very pure germanium detectors. The 

abundance, however, is very small. This type of measurement is difficult and rarely done in 

most laboratories, even today. SC&A does not believe it was possible to measure Th-232 this 

way in 1966–1969. Thus, it may not make sense to apply the minimum detectable activity 

(MDA) from the 1990s for 1966–1969. 

Thorium is, in general, measured through its daughters. The best daughter to measure is Ac-228 

because it comes before Rn-220. Rn-220 makes equilibrium assumptions more problematic. 

The measurement of Ac-228 requires the knowledge of equilibrium factors between Th-232, Ra228, and Ac-228. Another problem is that Ra-228 leaves the lung faster than Th-232; thus, the 

assessment of intakes based on measurements of Ac-228, or the other daughters, can be 

underestimated by a large and unknown amount (up to two orders of magnitude for Type M). 

In summary, in vivo bioassay monitoring results for thorium should be analyzed very 

carefully. Errors of two orders of magnitude can be made, depending on the material 

type, equilibrium assumptions, and time of measurement after intake. Urinalysis results 

should also require a very cautious analysis, including the influence of natural thorium in 

the diet. 

With respect to the added discussion on U-233 (including impurities in uranium), NIOSH 

may need to do more to check the reasonableness of the impurity ratio assumptions. 

Assuming U-233 is like U-234 is most likely reasonable, although the half-lives are 

different. 

4.1.1.4 Trivalent Actinides 

Bioassay for trivalent actinides (americium, curium, and californium) was not available until the 

mid-1960s. The exposure potential for these radionuclides predates the development of a 

bioassay method. For example, one of the first incidents at SRS in September 1954 involved a 

spread of contamination from an americium source (Nichols et al. 1954). Product and waste 

streams likely contained concentrations of Am-241. The lack of monitoring data from years 

prior to bioassay monitoring brings into question whether all intakes were captured. 

Section 4.1.2 of Revision 03 provides information on trivalent actinide monitoring, including 

minimum detectable activity for Am-241, Np-237, Cm-244, and Cf-252 since 1994. Gross alpha 

bioassay results prior to 1994 are used to assign dose from trivalent actinides. In the absence of 

specific information on the principal radionuclide, the dose reconstructor is told to assume 

Am-241 (Scalsky 2005, p. 65). This provides a methodology for reconstructing dose when 

monitoring data are available. Potential unmonitored exposures have not been considered. 

Furthermore, proposed Revision 04-E includes modifications. SC&A finds that the statement in 

proposed Revision 04-E (Scalsky 2006) on page 36 at the bottom and on page 37 at the top, 

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which indicates that the majority of exposure is in the neptunium cycle, does not appear to be 

justified by any data. According to the TBD, the exposure it is from plutonium and cerium. 

On page 74 (Scalsky 2006), the discussion of neptunium MDA includes apparently inconsistent 

statements. First, the MDA is cited as 0.035 dpm/liter (L), and from a “personal 

communication,” it is cited as 0.4 dpm/L. Later in the discussion, it is indicated that it is 

acceptable to use numbers less than 0.4 dpm/L, as reported. 

NIOSH has acknowledged that the history of potential exposure to neptunium and curium prior 

to the 1960s requires more research. It was further noted by NIOSH/ORAUT that the activity of 

these elements would have been less than in the 1960s and would have involved only a few 

workers associated with feasibility studies. The basis for assumptions on the neptunium cycle 

should be documented. Consideration also needs to be given to americium and californium. 

Numerous special radionuclides were handled at SRS, ranging in quantities from fractions of a 

gram to kilograms. Many of the sources produced were encapsulated and therefore posed 

primarily an external hazard. Some of the special radionuclides handled at SRS included 

Po-210, Co-60, Cf-252, Tm-170, Ir-192, Eu-152, and various isotopes of lanthanum (Reed et al. 

2002). Revision 03 of the SRS TBD gives inadequate or no consideration to potential exposures 

and missed dose from these radionuclides. It does not discuss the implementation of monitoring 

techniques for these radionuclides. 

Revision 04-E (Scalsky 2006) of the TBD adds the following information related to special 

radionuclides production and the High Neutron Flux program. 

Starting in 1955 the reactors were used to produce megacuries of 60Co for 

industrial and medical uses. From 1965 through 1970, C and K reactors were 

modified to be able to produce a very high flux of neutrons for production of 

unusual radionuclide sources, referred to as the High Flux program. Curies to 

megacuries of an estimated 150 different radionuclides were produced 

(Bebbington 1990, DuPont 1965, Gray 2006). These sources were encapsulated 

prior to irradiation and shipped offsite so potential for intake was minimal; the 

sources were inspected and loaded into casks under water to reduce the external 

exposure (Gray 2006). 

This statement seems to indicate that internal dose will not be calculated unless specific 

situations are noted in the dosimetry file. This will be further evaluated by SC&A at the 

finalization of Revision 04-E (Scalsky 2006). 

4.1.1.5 High Neutron Flux Programs 

The High Neutron Flux program involved the irradiation of targets to produce transplutonium 

radionuclides, such as curium and californium. Furthermore, the High Neutron Flux program 

also resulted in the production of high specific-activity Co-60. Revision 03 provides no 

guidance on how these radionuclides should be considered in dose reconstruction. 

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NIOSH added a paragraph to proposed Revision 04-E of the site profile indicating that sources in 

the High Neutron Flux program were sealed and shipped off site after removal from the reactors. 

The proposed TBD (Scalsky 2006, p. 25) states the following: 

...these sources were encapsulated prior to irradiation and shipped offsite so 

potential for intake was minimal:.... 

This seems to indicate that there was little potential for intake from the various “exotic” 

radionuclides made at SRS during the High Neutron Flux program. This does not seem to take 

into account incidents or exposure during encapsulation. As was the case in the Y-12 SEC 

evaluation, records may be available that would serve to characterize how these “other 

radionuclides” were handled, by whom, and in what amounts. The identification and retrieval of 

such records would shed light on this potential source of exposure. NIOSH should look more 

carefully at additional sources, such as monthly progress reports, “Bebbington 1990,” and “Gray 

2006,” which discuss such “exotics.” It is clear that there were neptunium, curium, and 

americium exposures (Du Pont 1965). 

4.1.2 Applicability to Revision 03 

Revision 03 includes a new Section 4.1.2 on bioassay, but Comment 1 remains substantially the 

same for the updated TBD. Several updates proposed for Revision 04-E (Scalsky 2006) of the 

TBD deal directly with this issue. This issue remains open pending review of Revision 04-E of 

the TBD, when it is issued, and ORAUT-OTIB-0053, Dose Reconstruction Considerations for 

Recycled Uranium Contaminants (ORAUT 2005). 

4.1.3 Summary of Overall NIOSH Response 

A table providing impurities to be included with intakes of uranium has been drafted for the next 

revision of Chapter 4 of the SRS TBD. This table will provide the six impurities with the largest 

impact on dose, with different impurities for depleted uranium or natural uranium and enriched 

uranium. The impurity values were obtained from Historical Generation and Flow of Recycled 

Uranium at the Savannah River Site (McCarty 2000), which relied heavily on Dose Contribution 

from Plutonium and Other Impurities in Uranium in Waste Streams from Savannah River Site 

Uranium Recovery Facilities (Crase and LaBone 2000). In general, the SRS impurities are lower 

than those used for Hanford. Neither set of impurities adds much to the internal organ doses. In 

addition, NIOSH is developing ORAUT-OTIB-0053 (ORAUT 2005) to provide generic 

guidance on RU. NIOSH indicated in its initial issue resolution response that this Oak Ridge 

Associated Universities Team Technical Information Bulletin (OTIB) would be released in 

December 2006. 

A brief description of the higher wt.% Pu-242 source term was added in the plutonium bioassay 

section. The effect of the higher wt.% Pu-242 was investigated for its impact on plutonium 

bioassay yield calculations; however, the high Pu-242 campaigns ended in 1967, long before 

alpha spectrometry was used for plutonium bioassay. The impact on yield would have been less 

than 10%, which is adequately accounted for in the overall geometric standard deviation of 3 

used for intake uncertainty. 

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Bioassay for americium, curium, and californium was in place during the mid-1960s. 

The first irradiation of targets to produce Co-60 appears to have occurred in August 1956. Those 

targets were encapsulated and after irradiation were shipped directly to Oak Ridge National 

Laboratory. Hence, there was little chance for intake at SRS. Work began on a urinalysis 

procedure for Co-60, Fe-59, Zn-65, and Cr-51 in November 1956, but the impetus appears to be 

normal reactor activation products, not special targets. This procedure appears not to have been 

used routinely until 1961; that is, its use was event driven. Whole-body counts starting in 

December 1960 do not list Co-60 as a frequently detected radionuclide despite its easy detection 

relative to other fission/activation products that were listed. The targets were encapsulated and 

Co-60 was not often seen in whole-body counts, leading to the conclusion that the Co-60 source 

program was not a source of unmonitored intakes. 

A description of the thorium/U-233 operations will be added to Section 4.1.2 of the 

Revision 04-E (Scalsky 2006) of the SRS TBD, along with some description of the bioassay and 

instructions for incorporating default thorium intakes for M-Area and canyon buildings workers. 

With regard to radionuclides from the High Neutron Flux program, whole-body counting would 

have detected intakes of most of the special radionuclides, but it is agreed that it would not have 

been appropriate for the beta-only or low-energy photon-emitting radionuclides, Tm-170 being 

most notable. The reason that specific in vitro bioassay analyses were not developed for Tm-170 

and like sources is that the preirradiated source material was in welded aluminum cans; after 

irradiation, the cans were placed directly into a shielded shipping cask under water. The sources 

were shipped to customers without being opened. The potential for intakes was low. 

NIOSH has added a paragraph to proposed Revision 04-E (Scalsky 2006) of the site profile 

indicating that sources in the High Neutron Flux program were sealed and shipped off site after 

removal from the reactors. Because the sources were encapsulated prior to irradiation and 

shipped offsite, NIOSH indicates there was minimal potential for intake. This does not seem to 

take into account incidents or exposure during encapsulation. As was the case in theY-12 SEC 

evaluation, records may be available that would serve to characterize how these “other 

radionuclides” were handled, by whom, and in what amounts. The identification and retrieval of 

such records would shed light on this potential source of exposure. 

4.1.4 Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Provide a timeline for the development and release of the RU OTIB (pending) 

•	 Provide intended approach for addressing special nuclides 

•	 Provide additional information on thorium and U-233 and the intended approach for 

addressing this issue 

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•	 Provide justification for the ability to link to the bioassay program and substantiate 

extrapolation back to the earliest timeframe 

•	 Provide information regarding the monitoring of uranium at SRS 

•	 Provide an intended approached for addressing potential intakes from the Pu-242 

program
 

•	 Conduct further research as indicated and appropriate regarding potential exposure to 

trivalent actinides 

SC&A Actions: 

• Provide a timeline for the review of the RU OTIB when it becomes available (pending) 

• Review discussions added to the revised site profile (Revision 04E) when available 

4.1.5	 Closure Status 

The assessment and guidance on dose assignment is incomplete pertaining to RU, Pu-242, 

transplutonium radionuclides, thorium, and U-233 in Revision 03 of the TBD. Section 4.1.2 on 

bioassay was added in Revision 03, but the issues related to these radionuclides have not been 

adequately resolved. Several updates proposed for Revision 04-E of the TBD deal directly with 

exposure from these radionuclides and bridge some of the gaps found in previous versions of the 

TBD. This draft revision will respond to several of the NIOSH actions items requested by the 

Work Group. It is recommended that this overall issue remain open pending review of yet-to-bereleased Revision 04-E (Scalsky 2006) of the TBD and ORAUT-OTIB-0053 (ORAUT 2005). 

4.2	 COMMENT 2: BETA/GAMMA ADJUSTMENT AND UNCERTAINTY 

FACTORS 

The assessment of the beta/gamma dosimeter adjustment factors and uncertainties within the 

TBD is incomplete, resulting in a probable underestimation of doses. The TBD needs to be more 

specific and complete in the following areas: 

•	 The calibration of dosimeters is often not representative of incident angles encountered in 

the field and, depending on exposure geometry, could result in an underestimation of the 

true exposure that is being measured. 

•	 The on-phantom correction factor of 1.119 may be too low for photon energies between 

30 and 250 kiloelectron volt (keV). 

•	 The TBD’s 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%. 

•	 Dosimeter adjustment factors for SRS are inconsistent with DOE complex-wide TIBs. 

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SC&A originally identified an initial lack of guidance pertaining to shallow doses, which has 

since been resolved. A minor deficiency of this TBD, particularly Section 5.0, is the absence of 

guidance pertaining to the interpretation of open window dose, shallow dose, 7 milligrams per 

square centimeter (mg/cm2

) dose, and/or skin dose. SC&A has reviewed ORAUT-OTIB-0017, 

Interpretation of Dosimetry Data for Assignment of Shallow Dose (Merwin 2005), which was 

released since SC&A’s original evaluation, and found that it satisfactorily addresses shallow 

doses. 

4.2.1 Issue Description 

4.2.1.1 Dose Adjustment Factors 

In the mid-1980s, SRS implemented changes in the calibration of dosimeters that replaced the 

previous Ra-226 source with a Cs-137 source and switched from in-air calibration of dosimeters 

to on-phantom. The overall change in recorded dose was assumed to require a correction factor 

of 1.119 for dosimeter readings prior to 1986 and 1.039 for dosimeter results during 1986 

(Taylor et al. 1995). These factors are recommended as dose adjustment factors (DAFs) in the 

TBD. These DAFs were generated for the purpose of comparing doses across timelines 

(i.e., normalizing recorded doses from these earlier periods in terms of 1995 dose assignment 

methods). The DAFs were developed by comparing thermoluminescent dosimeter (TLD) 

responses during calibrations under different conditions: air versus phantom, and Cs-137 versus 

Ra-226 as the gamma calibration source (Taylor et al. 1995, pp. 36–37). These adjustment 

factors were intended for comparative purposes only; the authors did not intend to use them for 

amending individual dose records (Taylor et al. 1995, p. 93). 

In Revision 03 of the TBD, NIOSH/ORAUT acknowledged that multiple factors may contribute 

to an under-response of a film dosimeter or TLD. However, they conclude that the dosimeters’ 

over-response at low photon energies was offset by under-responses caused by calibration 

methods, angular response, environmental factors, and other considerations. Furthermore, they 

indicate that the recorded dose for all types of dosimeters employed was, in fact, a reasonable 

estimate of the 1,000 mg/cm2

 deep dose. Since the over-response was energy dependent and was 

limited to the two-element film used from 1951–1959, SC&A did not agree that it would offset 

the many factors contributing to under-response. It appeared likely that doses recorded after 

1959 and not impacted by the over-response phenomenon would be significantly underestimated. 

Hine and Brownell (1956) have evaluated backscatter and concluded that it depends in a 

complex way on (1) the energy of the radiation, (2) the area of the field, and (3) thickness of the 

scattering medium. The percentage of backscatter may be as high as 50% for a large field, 

adequate thickness, and select photon energy. The data indicate that for photons with half-value 

layer between 0.6 millimeters Copper (mm Cu) and 1.0 mm Cu (or approximately 60–80 keV), 

the backscatter factor for a dosimeter worn on the upper torso of an adult could reach a value of 

about 1.5. Such a backscatter factor would apply to dose conversion factors (DCFs) with photon 

energies between 30 keV and 250 keV, which is commonly assumed for SRS workers. 

SC&A reviewed the available references relevant to beta/photon dosimetry at SRS. SC&A also 

reviewed Revision 03 and proposed Revision 04-E of the SRS TBD (Scalsky 2005, Scalsky 

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2006) and a comparison study of DOE external dosimetry (Alvarez 2003). The following 

concerns are noted regarding the use of these DAFs within the TBD: 

•	 Application to TLDs Over Time: While the methods and data used to derive the DAFs 

may be sound for the study’s intended purpose, the resulting factors do not necessarily 

represent comprehensive dose adjustment factors valid for the entire period to which they 

are being applied. TLDs were first used for official records in April 1970; it has not been 

shown that one DAF of 1.119 is appropriate for the entire time span from 1970 through 

1985. 

•	 Application to Film Badges: It is feasible that a detailed analysis might produce yearand location-specific DAFs in the 5–15% range for TLDs. However, the TBD has not 

addressed the validity of applying these DAFs for film badges. One DAF of 1.119 

(derived from TLD data) would not appear appropriate for film-based dosimetry used at 

SRS from 1952 through 1969. This issue is not addressed in Revision 03 or the draft 

Revision 04-E (Scalsky 2005, Scalsky 2006) of the SRS TBD. 

•	 Impact on Coworker Data: ORAUT-OTIB-0032, External Coworker Dosimetry Data for 

the Savannah River Site (Merwin 2006), lists the recommended coworker doses to be 

assigned to unmonitored or under-monitored workers at SRS. These data were taken 

from the Health Protection Annual Radiation Exposure History (HPAREH) database and 

adjusted by the correction factor of 1.119 prior to 1986 (Merwin 2006). Apparently, the 

1.039 adjustment factor for 1986 was not taken into account. The use of a single 

adjustment factor for film and TLD dosimeters from 1952 through 1985 will affect the 

dose assigned to unmonitored workers, as well as the monitored workers, and could result 

in an underestimate of their doses. 

•	 Limitations of DAFs: Section E.4.1.2 of the TBD (Scalsky 2005, Scalsky 2006) states 

that adjustments are needed to account for uncertainties associated with complex 

workplace radiation fields and exposure orientations. The DAFs of 1.119 and 1.039 are 

assumed to address this need. However, it is important to realize that a DAF mainly 

corrects for bias in the calibration system/processing and not for other variables and 

uncertainties. Section 4.2.1.2 of this report addresses these uncertainties. 

The uncertainty identified for adjustment factors to the beta/gamma dosimeter-measured dose 

does not entirely estimate uncertainty from all potential field exposure conditions. For this 

reason, other compensating sources of uncertainty must be considered, including laboratory 

uncertainty for measured dose, missed dose uncertainty as defined in OCAS-IG-001, External 

Dose Reconstruction Implementation Guideline (NIOSH 2002), and, for some workplaces, 

uncertainty for assigned neutron dose (measured, missed, and unmonitored), as well as the 

Interactive RadioEpidemiologic Program (IREP) method to calculate the target organ dose [and 

thus the probability of causation (POC)]. Based on the overall effect of the uncertainty 

incorporated in this analysis, the assigned organ dose used in compensability determination does 

not underestimate the actual dose. 

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4.2.1.2 Uncertainty Factors 

SC&A’s searches and reviews have not yielded further qualification for the uncertainty factors 

described in the TBD (Scalsky 2005, Scalsky 2006). The SRS documents that SC&A has been 

able to find do not deal specifically with uncertainties or adjustment factors, and SC&A could 

not locate the “3/29/04 guide” identified in NIOSH’s reference list. Additional information 

would be needed in order for SC&A to locate this document and evaluate its applicability to this 

issue. 

The revised TBD addresses recorded dose uncertainties in Section 5.3.5, Table 5.3.5-1 of 

Revision 03 (Scalsky 2005) and Section 5.3.5, Table 5-20 of Revision 04-E (Scalsky 2006). 

Table 5.3.5-1 (Revision 03) is reproduced below: 

Dosimeter SRS Laboratory 

uncertaintya 

Workplace uncertaintyb 

Reactor Plutonium Reprocessing 

Beta/gamma dosimeters 

 Two-element film Used 1951–1959 +/- 25% +/- 50% +/- 75% +/- 50% 

 Multi-element film Used 1960–1969 +/- 20% +/- 40% +/- 60% +/- 40% 

TLD Used 1970–present +/- 10% +/- 20% +/- 30% +/- 20% 

Neutron dosimeters 

NTA Used 1951–1970 +/- 50% +/- 100% (need to use another method) 

TLND Used 1971–present +/- 25% +/- 50% +/- 75% +/- 75% 

a. In relation to Hp(10) response of dosimeter. 

b. 95% confidence interval. 

These tables list laboratory uncertainties (excluding neutron track emulsion (NTA)) ranging from 

10–25% and workplace uncertainties ranging from 20–75%, but they do not state what variables 

are addressed (i.e., energy response, mixed fields, geometry). There is no further quantitative 

discussion of laboratory uncertainties (e.g., calibration, chemical processing, reading); 

radiological uncertainties (changes in energy spectra, geometry of exposure, and other field 

variables); and environmental uncertainties (e.g., fading, moisture, light, temperature, chemical 

exposures). 

The TBD continues to recommend a generic standard deviation of 30% for best estimate dose 

reconstruction or another appropriate value from Section 5 [Revision 03, Section 5.7.2, p. 117 

(Scalsky 2005); Revision 04-E, Section 5.10.6, p. 121 (Scalsky 2006)]. It would appear that the 

combined uncertainties from the large range of uncertainties listed in Table 5.3.5-1 and Table 5­

20 would be more in the range of 40–50% under actual working conditions, especially during the 

earlier years of film dosimetry, such as before 1971. For example, combining the laboratory and 

workplace uncertainties using simple quadratic calculations gives a range of 22% to 79% 

(excluding NTA), with an average of around 55%. 

If an uncertainty value of 30% is to be retained in the TBD, detailed mathematical derivation of 

this value should be provided as it applies to SRS. Additionally, the statement that “measured 

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doses are to be treated as a normal distribution with a standard deviation of 30% (or other 

appropriate value that may be provided in Section 5)” lacks clarity and guidance for the dose 

reconstructor. Does “other appropriate value” refer to a combination of uncertainties listed in 

Table 5.3.5-1 (Revision 03, Scalsky 2005) or Table 5-20 (Revision 04-E, Scalsky 2006)? If so, 

how are they to be combined? The uncertainty factors provided in Section 5.3.5 of the TBD are 

similar to those found in some of the documents from NIOSH’s reference list (Attachment B of 8 

November 2006), but a final combination value or mechanism is needed in order for dose 

reconstructors to consistently apply all of the relevant uncertainties to the dose of record. 

4.2.1.3 Shallow Dose 

SC&A originally identified an initial lack of guidance pertaining to shallow doses, which has 

since been resolved. A minor deficiency of this TBD is the absence of guidance pertaining to the 

interpretation of open window dose, shallow dose, 7 mg/cm2 dose, and/or skin dose. Although 

these terms are defined in the glossary and mentioned in the Executive Summary of Revision 03 

(see page 17, which states, “... Section 5 presents the occupational dosimeter program for 

measuring skin and whole-body doses to workers” [Emphasis added.]), there is neither a 

discussion of shallow dose interpretation nor a reference to ORAUT-OTIB-0017 (Merwin 2005). 

SC&A reviewed ORAUT-OTIB-0017 (Merwin 2005), which was released since SC&A’s 

original evaluation, and found that it satisfactorily addresses shallow doses. 

4.2.2 Applicability to Revision 03 

Comments regarding adjustment factors are still applicable. ORAUT-OTIB-0017 (Merwin 

2005) provides guidance for shallow dose assignment; hence, this is no longer an issue. 

4.2.3 Summary of NIOSH Response 

The information in the TBD regarding correction factors for beta/gamma dosimetry is based on 

SRS historical evaluations of dosimeter performance. NIOSH provided the following references 

as the historic basis for dosimeter performance: 

•	 Atomic Energy Commission, 1955, “Intercomparison of Film Badge Interpretations,” 

Isotopics, Volume 2, number 5, pp. 8–23. 

•	 Brackenbush, L.W., G.W.R. Endres, J. M. Selby, and E.J. Vallario, 1980, “Personnel 

Neutron Dosimetry at Department of Energy Facilities,” PNL-3213, Pacific Northwest 

Laboratory, Richland, Washington. 99352. 

•	 Brackenbush, L.W., K.L. Soldat, D. L. Haggard, L. G. Faust and P. L. Tomeraasen, 1987, 

“Neutron Dose and Energy Spectra Measurements at Savannah River Plant,” PNL-6301, 

Pacific Northwest Laboratory, Richland, Washington. 99352. 

•	 Brodsky, A., and R.L. Kathren, 1963, “Accuracy and Sensitivity of Film Measurements 

of Gamma Radiation—Part I: Comparison of Multiple-Film and Single-Quarterly-Film 

Measurements of Gamma Dose at Several Environmental Conditions,” Health Phys 

9(4):453–461. 

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•	 Brodsky, A., and R.L. Kathren, 1963, “Accuracy and Sensitivity of Film Measurements 

of Gamma Radiation—Part II: Limits of Sensitivity and Precision,” Health Phys 

9(5):463–471. 

•	 Brodsky, A., A.A. Spritzer, F.E. Feagin, F.J. Bradley, G. Karches and H.I. Mandelberg, 

1963, “Accuracy and Sensitivity of Film Measurements of Gamma Radiation—Part IV: 

Intrinsic and Extrinsic Errors,” Health Phys 11(10):1071–1082. 

•	 Gorson, R.O., N. Suntharalingam, and J.W. Thomas, 1963, “Results of a Film-Badge 

Reliability Study,” Presented at the 49th Annual Meeting of the Radiological Society of 

North America, Chicago, Illinois. November 17–22, 1963. 

•	 Hoy, J.E., 1972, “Personnel Albedo Neutron Dosimeter with Thermoluminescent 7Li and 

6Li,” DP-1277, E.I. du Pont de Nemours and Company, Inc., Savannah River Plant, 

Aiken, South Carolina. 

•	 Hoy, J.E., 1980, “Thermoluminescent Dosimeters for Personnel Neutron Monitoring,” 

DPST-70-533, E.I. du Pont de Nemours and Company, Inc., Savannah River Plant, 

Aiken, South Carolina. 

•	 Morgan, K.Z., 1961, “Dosimetry Requirements for Protection from Ionizing Radiation,” 

Selected Topics in Radiation Dosimetry, Proceedings of the Symposium on Selected 

Topics in Radiation Dosimetry, Sponsored by and Held in Vienna 7–11 June 1960, 

International Atomic Energy Agency, Vienna, Austria, pp. 3–23. 

•	 Parker, H.M., 1945, “Comparison of Badge Film Readings at the Metallurgical 

Laboratories, Clinton Laboratories and the Hanford Engineering Works,” 7-3090, 

Hanford Atomic Products Operation, Richland, Washington. (SRS used Clinton 

dosimetry services and then implemented systems similar to other laboratories.) 

•	 Plato, P., 1978, “Testing and Evaluating Personal Dosimetry Services in 1976,” Health 

Phys 34(3):219–223. 

•	 Savannah River Site, 1993, “External Dosimetry Technical Basis Manual,” WSRC-IM92-101, Westinghouse Savannah River Company, Aiken, South Carolina. 

•	 Taylor, G.A., K.W. Crase, T.R. LaBone, and W.H. Wilkie, 1995, “A History of 

Personnel Radiation Dosimetry at the Savannah River Site,” WSRC-RP-95-234, 

Westinghouse Savannah River Company, Aiken, South Carolina. 

•	 Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis and P. Andreo, 2002, “Study of a 

Selection of 10 Historical Types of Dosemeter: Variation of the Response to Hp(10) with 

Photon Energy and Geometry of Exposure,” Radiat Prot Dos, 102(2):101–113. (Includes 

SRS Panasonic 802 dosimeter.) 

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•	 Vallario, E.J., D.E. Hankins, and C.M. Unruh, 1969, “AEC Workshop on Personnel 

Neutron Dosimetry,” BNWL-1340, Battelle, Pacific Northwest Laboratory, Richland, 

Washington. 

•	 Vallario, E.J., D.E. Hankins, and C.M. Unruh, 1971, “Second AEC Workshop on 

Personnel Neutron Dosimetry,” BNWL-1616, Battelle, Pacific Northwest Laboratory, 

Richland, Washington. 

•	 Wilson, R.H., J.J. Fix, W.V. Baumgartner, and L.L. Nichols, 1990, “Description and 

Evaluation of the Hanford Personnel Dosimeter Program from 1944 Through 1989,” 

PNL-7447, Pacific Northwest Laboratory, Richland, Washington. (See Section 5.3 on 

intercomparison programs.) 

The TBD excludes beta and nonpenetrating doses for the extremity, skin, gonads, and breast 

exposure in Section 5.1. 

ORAUT-OTIB-0017 (Merwin 2005) provides guidance for assessing the shallow dose and 

nonuniform exposures. This guidance was present in the draft TBD (Revision 04-E, Scalsky 

2006) at the time of this review. 

4.2.4 Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Provide references on adjustment factors used. (NIOSH supplied a list of references on 

November 8, 2006) 

SC&A Actions: 

•	 Compare the uncertainty factors in the TBD against relevant site-specific workbooks as 

well as available dose reconstruction instructions (e.g., March 29, 2004, guide). 

4.2.5 Closure Status 

The issues associated with correction factors and uncertainties have not been satisfactorily 

resolved. It has not been demonstrated that the application of a DAF of 1.119 or 1.039 for both 

TLDs and film for 1952–1986, and an uncertainty of 30% without full consideration of 

laboratory, radiological, and environmental factors, is claimant favorable. The dosimeter 

calibration is based on an incident angle of zero degrees, which underestimates the actual field 

dose where incident angle is greater than zero. The correction factor applied to recorded 

dosimeter results is too low for photon energies from 30 to 250 keV, which is the default photon 

energy used. 

SC&A recommended the following specific actions to achieve closure: 

• Provide more detailed period/location film-specific DAFs for the period 1952–1970. 

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•	 Provide period/location TLD-specific DAFs for the period 1971–1986. 

•	 Assess the impact of new DAFs on coworker data. 

•	 Account for differences in incident angles between calibration and field use. 

•	 Account for photon energies between 30 and 250 keV (the default photon energy used in 

calibration). 

•	 Clarify the basis for applying a generic 30% uncertainty factor and/or provide clear 

instructions for applying other appropriate uncertainty factors. 

Revision 03 does not provide addition information to resolve this issue. The recent draft of 

Revision 04-E (Scalsky 2006), which has not yet been officially issued, contains a few changes 

to external dose reconstruction, but these additions do not satisfactorily address the original 

issues; therefore, the issues are still applicable. None of the SRS site-specific workbooks and 

guides that SC&A has been able to locate provide further qualification for using the DAFs 

as recommended in the TBD (Scalsky 2006). The SRS documents that SC&A has been able to 

find do not deal specifically with uncertainties or adjustment factors, and SC&A needs more 

information on the title or other bibliographic data to locate the guide of 3/29/04 (as listed in 

the NIOSH response) and evaluate its applicability to this issue. SC&A recommends that this 

issue remain open for these reasons. SC&A concurs that ORAUT-OTIB-0017 (Merwin 2005) 

provides suitable guidance for the assignment of shallow dose. 

4.3 COMMENT 3: INADEQUATE NEUTRON-TO-PHOTON RATIOS 

Questions remain unanswered regarding the validity of neutron-to-photon ratios at SRS. The 

geometric mean and standard deviation that describe the post-1971 neutron-to-photon ratio are 

not technically sound or likely to be claimant favorable for a large number of claimants. The 

TLND-recorded neutron doses between 1971 and 1995, as well as the pre-1971 neutron doses 

(derived from neutron-to-photon ratios), suffer from a high degree of uncertainty. The use of the 

95th percentile value for the TLND neutron dose of record is recommended. 

4.3.1 Issue Description 

Neutron dosimetry is considerably more complex and difficult to assess than beta/photon 

dosimetry. Difficulties in assessing neutron dose are principally the result of design limitations 

of past dosimeters used at SRS, and the highly variable and complex neutron spectra that 

workers may have encountered. The four main areas at SRS with potential for neutron exposure 

include the plutonium facilities in the 200 Area; the Calibration Facility (736-A) and the Cf-252 

Facility (773-A) in the 700 Area; reactors in the 100 Area; and Building 321 (plutoniumaluminum alloys) in the 300 Area. These facilities differ not only in neutron energy spectra but 

also in terms of their neutron-to-photon dose rate ratios. The significance of the latter is highly 

relevant to the time period at SRS when neutron exposure was assessed by neutron track 

emulsion (NTA) film. 

Neutron dosimeters used to monitor individual workers at SRS involved three different designs. 

The first involved the NTA Type A film dosimeter, which was used from August 3, 1953, 

through the end of 1970. This dosimeter relied on the interaction of neutrons with sensitive 

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elements of the film to produce visible tracks. When manually counted, the number of tracks per 

film provides an estimate of the total neutron fluence to which a worker was exposed and from 

which an estimate of neutron dose is derived. 

Due to the insensitivity of NTA film to neutrons with energies below 500 keV (or even 1 

megaelectron volt (MeV), as reported by others), as well as other factors contributing to the 

dosimeter’s uncertainty, NIOSH/ORAUT concluded that NTA monitoring data used from 1953 

through the end of 1970 were insufficiently reliable and therefore could not be used for dose 

reconstruction. It was concluded that a suitable substitute for NTA neutron data was the use of 

facility-specific neutron-to-photon ratio data. 

The SRS TLD neutron dosimeter was introduced on January 1, 1971, and was used until 

December 31, 1994. The TLND employed two polyethylene spheres covered with a layer of 

cadmium. Proper interpretation of this dosimeter requires matching the neutron energy spectrum 

of the calibration source with that of the workplace spectrum. 

On January 1, 1995, SRS began using the commercial Panasonic neutron TLD, which detects 

albedo neutrons. Albedo neutrons are those reflected backwards out of the worker’s body into 

the TLD’s phosphor, where the neutron interacts with Li-6 to give an alpha particle and tritium 

(i.e., n + Li-6 → α + H-3). A combination of Li-7 and Li-6 phosphors, along with multiple 

filters, and an empirically derived algorithm allows this dosimeter to quantify exposure to betas, 

low-energy photons, high-energy photons, and neutrons. 

In order to assign neutron doses to workers who had been monitored by means of NTA film prior 

to 1971, the surrogate use of the neutron-to-photon ratio method required NIOSH/ORAUT to 

assess the neutron-to-photon dose rate ratios for each major location that posed the potential for 

neutron exposure between 1953 and the end of 1970. [One location where the use of NTA film 

dosimeters was considered useable is the Fuel Fabrication Area (321 M Area).] NIOSH/ORAUT 

employed empirical, location-specific neutron-to-photon ratios that were found to represent a 

lognormal distribution. These data could then be used to estimate neutron exposures on the basis 

of (1) recorded photon doses and (2) photon doses recorded as zero (i.e., missed photon doses). 

Starting in January 1971, neutron doses were monitored and recorded by means of the SRS Hoy 

TLND and the Panasonic TLD. NIOSH/ORAUT regards monitoring data for these dosimeters 

as “reasonably accurate” and therefore, useable for dose reconstruction, but not without 

“adjustment.” Since 1971, neutron doses recorded by TLDs were based on neutron quality 

factors in National Council on Radiation Protection and Measurement (NCRP) 38, Protection 

Against Neutron Radiation (NCRP 1971), which assigned specific values to discrete neutron 

energy intervals. Neutron quality factors defined in NCRP 38, however, have been updated by 

weighting factors in International Commission on Radiation Protection (ICRP) 60, 

Recommendations of the International Commission on Radiological Protection (ICRP 1990). In 

compliance with 42 CFR Part 82, NIOSH/ORAUT evaluated the neutron energy spectra at each 

of the major locations and provided corresponding location-specific neutron correction factors 

that account for revised neutron quality factors for post-1971 recorded neutron doses. 

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4.3.1.1 Neutron-to-Photon Ratio Method 

The TBD prescribes two very different protocols for neutron dose reconstruction that correspond 

to pre- and post-1971 time periods. SC&A’s review comments are therefore directed to each of 

these methods separately. 

SC&A reviewed the scientific literature regarding the use of NTA film dosimeters and agrees 

with the decision not to use NTA film data in dose reconstruction. SC&A further agrees with the 

use of the neutron-to-photon ratio method as a reasonable surrogate, but only on a conditional 

basis, as explained below. 

Of concern are the limited data that were used and the interpretation of such data for defining 

location-specific neutron-to-photon ratios. Table 4-3 below summarizes surrogate post-1971 

data that are to be used for neutron dose reconstruction prior to 1971. Values in bold indicate 

changes proposed for Revision 04-E (Scalsky 2006). 

Table 4-3. Neutron-to-Photon Ratio Values Used as Surrogate Data for 

NTA Film Dosimeters 

Areas/Process 

Neutron/Photon Ratio Neutron/Photon Ratio 

Avg. Range Geometric 

Mean 

Geometric 

Standard 

Deviation 

95th % 

100 Area—Reactors 0.26 (0.05–0.62) 0.18 2.52 0.82 

Plutonium Production: 

- HB-Line 

- FB-Line 

0.52 

1.29 

(0.09–1.23) 

(0.05–3.10) 

0.91 

1.0 

0.36 

1.0 

2.84 

2.0 

2.52 

2.0 

5.05 

3.1 

1.65 

3.1 

Radionuclide Production 

and Calibration 0.85 (0.10–3.83) 0.62 2.29 2.41 

Source: Scalsky 2005 

Table 4-2 shows that not only are there large differences in neutron-to-photon ratios among the 

four general areas, but even larger differences exist within a given area, as indicated by the wide 

range of ratio values. For example, at the FB-Line, observed ratio values, which range from a 

low of 0.05 to a high of 3.1, differ 62-fold. The observed wide range of neutron-to-photon ratios 

is clearly the aggregate of the following independent uncertainties of the post-1971 TLND 

neutron dosimeter: 

•	 The uncertainty of the post-1971 TLD photon dosimeter 

•	 The variability of the neutron-to-photon ratios among locations within a given area, such 

as the FB-Line 

In addition to these uncertainties are two more uncertainties that contribute to the actual pre-1971 

neutron dose. The third uncertainty is the pre-1971 photon dose (which must be multiplied with 

the post-1971 neutron-to-photon ratio), and the fourth uncertainty is the unfounded assumption 

that a post-1971 neutron-to-photon ratio at any of the four general areas is representative of the 

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pre-1971 neutron-to-photon ratios. This assumption would only hold true if all processes, 

production quantities, engineering controls, radiological practices, and other factors during the 

assessed post-1971 era were, in fact, identical/comparable to those that existed between 1953 and 

1970. Table 4-4 summarizes the uncertainties that collectively define the overall uncertainty of 

pre-1971 neutron doses that are derived by the photon-to-neutron ratio method. 

Table 4-4. Uncertainties Contributing to the Derivation of Neutron Dose by the 

Neutron-to-Photon Ratio Method 

Source of Uncertainty Workplace Uncertainty 

(1) Pre-1971 photon dose: 

­ Two-element film ±50% to ±75% (a) 

- Multi-element film ±40% to ±60% (a) 

(2) Post-1971 TLND dose ±50% to ±75% (a) 

(3) Post-1971 photon dose ±20% to ±30% (a) 

(4) Neutron-to-photon ratio 

 By locations within area Unknown 

(5) Neutron-to-photon ratio before 1971 

Versus measured neutron-to-photon

 Ratios post-1971 

Unknown 

(a) Source: Table 5.3.5-1 in ORAUT-TKBS-0003 

SC&A concludes that the surrogate use of the neutron-to-photon ratio method encompasses three 

large/quantifiable and two nonquantifiable uncertainties. Together, these uncertainties preclude 

the use of guidance, as given in Section E.4.1.6 of Attachment E of the TBD, which states the 

following: 

Prior to 1971, . . . using a ratio of the potential neutron dose to the measured 

photon dose is done as a claimant-favorable option to reconstruct an individual 

worker neutron dose . . . As can be determined from [Table] E-9, the 

recommended method to apply the ratio is as a lognormal distribution using the 

geometric mean and geometric standard deviation. [Emphasis added.] 

SC&A believes that the use of the geometric mean and geometric standard deviation that 

describe the post-1971 neutron-to-photon ratio is neither technically defensible nor likely to be 

claimant favorable for a large fraction of potential claimants. A claimant-favorable alternative is 

to use the 95th percentile neutron-to-photon ratio as a point estimate for all claimants regardless 

of compensability of the claim. 

4.3.1.2 Performance Characteristics of the TLND 

Closely linked to the issues identified above is SC&A’s second concern about the use of the 

TLND in its other role as the neutron dosimeter of record between 1971 and 1995. In part, 

Section 5.3.4.1.2 of the TBD explains the decision to accept the TLND data: 

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Trends in the annual SRS and Hanford neutron collective dose (Taylor et al. 

1995; Buschborn and Gilbert 1993, respectively), normalized to the annual 

plutonium production (DOE 1996), are illustrated in Figure 5.3.4.2-2. It is 

evident in this figure that the collective neutron dose was under-recorded prior 

to implementation on 1 January 1971 of the SRS TLND and 1 January 1972 

for the Hanford TLD. The extent of the under-estimate is difficult to estimate. 

SRS and Hanford showed a significant increase in the ratio of the annual 

collective neutron dose to the annual plutonium production when the TLD 

neutron dosimeters were implemented. [Emphasis added.] 

Figure 5.3.4.2-2 referenced above is reproduced below as Figure 4.1. Data shown in Figure 4.1 

do not support either statement emphasized in the quotation above: 

•	 At both SRS and Hanford, the rise in collective dose (supposedly standardized to 

plutonium production) began well before the advent of the TLND; and, in both cases, the 

standardized collective dose dropped precipitously after the implementation of the TLND 

with subsequent fluctuations. 

•	 Because the collective neutron doses were standardized (i.e., defined in person-rem per 

unit quantity of plutonium), the observed oscillations clearly indicate that the collective 

dose is not correlated with or linked to plutonium production but may very well be the 

result of variations in neutron fields that surround work conditions in a given area and the 

variable response of the TLND. 

Figure 4-1. Trends in SRS and Hanford Collective Neutron Dose Normalized to 

Plutonium Production
 

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Section 5.3.5 of the TBD discusses the uncertainty of the TLND’s response and includes the 

following statements: 

As reported in the PNNL report, measurements with the TEPC, multisphere 

system and 3HE spectrometer were in general agreement. The TLND agreed 

within about 30% for most measurement locations along the plutonium 

production lines and storage areas. The TLND was within a factor of 3 (i.e., 0.3 

to 3) for the extremes in neutron energy spectra encountered at the K-reactor 

door (i.e., highly thermalized field) and for a californium shipping cast (i.e., 

where most lower energy neutrons had been removed). Over long time periods, 

workers would generally be expected to be involved in several different exposure 

profiles that will serve to minimize the extremes identified. These results are 

indicative of the technical difficulties to accurately measure neutron dose in the 

workplace. Table 5.3.5-2 presents a summary of common workplace neutron 

dosimeter performance characteristics . . . Measurements of TLND performance 

at SRS in 1987 (Brackenbush et al. 1987) indicate that the SRS measured neutron 

dose with the TLND (beginning 1 January 1971) is reasonably correct. For dose 

reconstruction under EEOICPA a claimant favorable standard error estimate of 

50% should be made for neutron dosimetry between 1971 and 1985. 

Based on data provided above, the TBD provides no compelling evidence to suggest that the 

TLND dosimeter offered significant improvements over NTA film. From statements made in 

Section 5.3.5 of the TBD, it is also unclear whether the recommended “claimant-favorable” 

standard error of ±50% for the TLND represents a time-average value, as stated above (i.e., “... 

over long time periods, workers would generally be expected to be involved in several different 

exposure profiles that will serve to minimize the extremes identified”). 

In brief, this suggests that both the TLND-recorded neutron doses between 1971 and 1995 and 

the pre-1971 neutron doses (derived by neutron-to-photon ratios) suffer from a high degree of 

uncertainty and must be viewed with caution. SC&A recommends the use of a 95th percentile 

value for the TLND neutron dose of record. 

Specific concerns regarding the neutron-to-photon ratios at SRS are applicable to Revision 03 

and proposed Revision 04-E (Scalsky 2006) of the TBD; therefore, SC&A recommends the use 

of the 95th percentile neutron-to-photon value for all SRS neutron dose reconstruction cases. 

SC&A’s evaluation of Revision 03 (Scalsky 2005) did not reveal any changes related to this 

issue. 

Proposed Revision 04-E of the TBD (Scalsky 2006) provides some changes and clarifications to 

the applications of neutron-to-photon values for dose reconstruction. The site profile 

recommends that in likely noncompensible cases, the 95th percentile neutron-to-photon dose ratio 

be applied to the recorded dose. For plutonium facilities, it sets the neutron-to-photon geometric 

mean equal to 1.0, with the geometric standard deviation equal to 2.0 and the 95th percentile 

equal to 3.1, and for reactors it sets the neutron-to-photon geometric mean equal to 0.18, with the 

geometric standard deviation equal to 2.52 and the 95th percentile equal to 0.82. However, there 

are still some areas of concern in view of the following facts: 

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•	 Table 5-14, page 102, of the TBD (Revision 04-E) (Scalsky 2006) lists the average 

neutron-to-photon value for the HB-Line as 1.29, with a range up to 3.10. 

•	 Section 5.3.4.2.3.2 of the TBD (Revisions 03 and 04-E) states that Brackenbush et al. 

(1987) reported the neutron-to-photon dose rate value for the F-Areas as 2. 

•	 According to the TBD (Revisions 03 and 04-E) and NIOSH’s original matrix 

response, likely noncompensable cases will be assigned the 95th percentile, but the 

likely compensable cases will be assigned the geometric mean, and for best estimates 

it is only stated that the dose reconstructor provides a claimant-favorable analysis of 

dose. 

In view of the first two points listed above, a neutron-to-photon value of 1.0 would appear to be 

an average value rather than a claimant-favorable value. Therefore, as per SC&A’s original 

comment and considering the third point listed above, it is recommended that the 95th percentile 

neutron-to-photon values be used in all SRS dose reconstruction cases, not in just the likely 

noncompensable cases. 

4.3.2 Applicability to Revision 03 

The application of geometric mean and 95% values to dose reconstruction is a programmatic 

issue. Specifics regarding the neutron-to-photon ratios at SRS are applicable to the current 

revision of the TBD. 

4.3.3 Summary of Overall NIOSH Response 

Extensive and detailed claimant data are now available that could be examined to verify the 

recommended neutron-to-photon dose ratio statistical parameters contained in the site profile. 

Additional claimants were selected to expand the analysis to more workers, but determining the 

actual work location of SRS employees proved difficult. NIOSH/ORAUT discussed this with 

SRS Health Physics, and they considered the approach too uncertain to provide confidence in the 

analyses. 

The site profile recommends applying the 95th percentile neutron-to-photon dose ratio to the 

recorded and missed photon dose for workers routinely present in work areas prior to 1972 with 

a potential for neutron exposure. This is applied to all likely noncompensable claims. This value 

is based on the SRS post-1971 TLND measurements, as well as the results of Pacific Northwest 

National Laboratory (PNNL) dose measurements in SRS facilities and PNNL dose 

measurements at Hanford, because of similar workplace source terms and activities. The 

magnitude of the recommended neutron-to-photon dose ratios is consistent with the results of an 

Atomic Energy Commission technical investigation of neutron doses (using neutron-to-photon 

dose ratios) in 1972 at Hanford. For likely compensable claims, the geometric mean value of the 

neutron-to-photon dose ratio is applied and, if necessary, a Monte Carlo analysis performed, 

taking into consideration the 95th percentile value as part of a lognormal distribution. The 

overall assessment of the neutron dose component in a best estimate dose reconstruction for a 

SRS claimant provides a claimant-favorable analysis of the neutron organ dose. 

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4.3.4	 Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

• Provide a detailed description of neutron-to-photon ratio methods at SRS. 

4.3.5	 Closure Status 

The TBD prescribes two very different protocols for neutron dose reconstruction that correspond 

to pre- and post-1971 time periods. Prior to 1971, the uncertainty factors associated with the 

neutron-to-photon ratio are neither technically defensible nor likely to be claimant favorable. 

The TBD provides no compelling evidence to suggest that the TLND dosimeter offered 

significant improvements over NTA film. In brief, this suggests that both the TLND recorded 

neutron doses between 1971 and 1995 and the pre-1971 neutron doses (derived by neutron-tophoton ratios) suffer from a high degree of uncertainty and must be viewed with caution. 

SC&A’s evaluation of Revision 03 (Scalsky 2005) did not reveal any changes concerning this 

issue since the review of Revision 02. SC&A believes that the use of the geometric mean and 

geometric standard deviation that describe the post-1971 neutron-to-photon ratio is neither 

technically defensible nor likely to be claimant favorable for a large fraction of potential 

claimants. Proposed Revision 04-E of the TBD (Scalsky 2006) provides some changes and 

clarifications to the applications of neutron-to-photon values for dose reconstruction. The site 

profile recommends that, in likely noncompensible cases, the 95th percentile neutron-to-photon 

dose ratio be applied to the recorded dose. SC&A recommends the 95th percentile neutron-tophoton values be used in all SRS dose reconstruction cases, not in just the likely 

noncompensable cases. No Work Group action is currently pending, but it is recommended that 

this issue remain open pending the release of proposed Revision 04-E (Scalsky 2006) of the SRS 

TBD. 

4.4	 COMMENT 4: ADEQUACY OF THE F- AND H-AREA TANK FARM 

CHARACTERIZATION 

The adequacy of the F- and H-Area Tank Farm characterization in the TBD is questionable for 

use as dose reconstruction guidance. Data evaluation appears to be incomplete with regard to 

exposure conditions and uncertainty. This is particularly true for early periods of operation, 

where primary records involving key operations and incidents are lacking. Moreover, the TBD 

provides no references for the tank farm discussion and includes no analysis indicating how the 

conclusions were reached. 

4.4.1	 Issue Description 

The TBD guidance needs to be more specific and complete in the following areas: 

• Radionuclide lists are incomplete for both internal and external radiation. 

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•	 Early worker incident and contamination records may be seriously incomplete. 

•	 Raw data on incidents and high-radiation areas indicate that geometry of exposure may 

be a problem. 

•	 The potential for internal and external exposure to unmonitored workers in areas not 

designated as radiological control areas needs to be investigated. 

•	 The completeness and adequacy of tank farm data used in the TBD are in question. 

In this section, the term “tank farm workers” refers to all personnel who performed work around 

tanks in the F- and H-Area Tank Farms. 

4.4.1.1 Radionuclide Lists 

The TBD (Scalsky 2005, p. 31) gives the radionuclide lists for the F- and H-Area Tank Farms as 

follows: 

Internal exposure. The majority of the annual internal effective dose equivalent 

in the F Area combined waste tank is delivered by 90Sr, 144Ce, and 244Cm. The 

majority of the annual internal effective dose equivalent in the H Area combined 

waste tank is delivered by 90Sr, 144Ce, and 238Pu. 

External exposure. The majority of the external dose in the F Area Combined 

Tank Waste is delivered by 90Sr, 144Ce, 137Cs, and 106Ru. The majority of the 

external dose in the H Area Combined Tank Waste is delivered by 90Sr, 144Ce, and

238Pu. 

Discrepancies exist between radionuclides for the tank farms on page 31 as compared to those in 

Table A-14. Revision 03, Table A-14, includes Pu-241 and Am-241, both of which are listed as 

“[s]ignificant to external exposure.” Yet neither radionuclide appears in the list in the main text 

of the TBD. Attachment A of the TBD extracts source term data from ESH-HPT-960197, 

Facility Description (LaBone 1996), but does not discuss the basis and method of determination 

of key internal and external radionuclides in this document. NIOSH determined which 

radionuclides delivered “the majority” of the internal and external doses in the respective areas to 

the exclusion of others present in those areas. SC&A finds that the lists are incomplete. 

Revision 03 does not include Cs-137 and Ru-106 as radionuclides of concern for internal dose. 

As abundant fission products, Cs-137 and Ru-106 are both of concern for internal exposure and 

are readily soluble in liquid. It is unclear why they are not included in the internal exposure list 

of radionuclides for both the F- and H-Area Tank Farms, despite evidence of their importance. 

For example, a body burden of 2% of the maximum permissible limit of 30 microcuries 

(i.e., 600 nCi) of Cs-137 was estimated for a mechanic in the H-Area Tank Farm who was 

accidentally exposed to high-level waste on February 28, 1974. This is higher than all but one of 

the high-five Cs-137 intakes listed for SRS in ORAUT-OTIB-0001 (Brackett 2003), Table 1. 

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Similarly, the tank farm data bank contains records of internal Ru-106 exposure. Furthermore, it 

is unclear why Ru-106 is not listed as a radionuclide of importance for external exposure in the 

H-Area Tank Farm, since that set of tanks also contains fission products. Rh-106, the short-lived 

decay product of Ru-106, is a gamma emitter. Both Ru-106 and Rh-106 are also sources of beta 

radiation. Therefore, Ru-106 (including its decay product, Rh-106) should have been flagged as 

important to internal and external exposure in the tank farm. Proposed Revision 04-E (Scalsky 

2006) includes Cs-137 and Ru-106 in Table A-14, resolving some of the issue. 

The internal radionuclide list is incomplete in other ways; for example, Tc-99 is not included. A 

number of other radionuclides, such as Zr-95 (and its decay product Nb-95), should also be 

evaluated for inclusion in the list of radionuclides of concern in both the F- and H-Area Tank 

Farms. Finally, the tank farm radionuclide list does not include several radionuclides that were 

produced, processed, or used as target material, specifically Th-232, Np-237, Pu-242, and U-233. 

Since the TBD presents no analysis regarding the tank farm radionuclide lists, it is unclear 

whether these radionuclides were evaluated for inclusion and then excluded because they did not 

contribute significant dose, or whether they were simply omitted. In the case of Th-232, Np-237, 

and U-233, the TBD discusses their use but does not include them in the tank farm radionuclide 

list for reasons that are not explained. If these radionuclides have been evaluated, the analysis 

should be presented. If not, they should be evaluated. NIOSH/ORAUT also needs to be aware 

of the differences in the constituents of the tanks based on the processes that fed them. 

4.4.1.2 Early Tank Farm Workers 

The tank farm data bank is incomplete. The F- and H-Area Tank Farm data bank entry of 

August 24, 1965, states the following: 

Prior to 1965, information on instrument failure, pump failure, leaks in the waste 

tank system are not recorded unless the individual occurrence is of particular 

interest. (Makhijani et al. 1986) 

The tank farm data bank did not identify any criteria by which an occurrence would be judged to 

be “of particular interest.” But it is clear that the data bank is incomplete in a number of 

different ways. For example, the data bank contains no entry that explicitly shows the amount of 

worker exposure prior to 1960, although there are many entries after that date. The change in the 

frequency of entries per year in the tank farm data bank is another indication that the vast 

majority of incidents, maintenance problems, cleanup activities, and similar events associated 

with the tank farms were not recorded during the 1950s, 1960s, and at least part of the 1970s. 

Table 4-5, reproduced from Makhijani et al. 1986, p. 30, shows the increasing frequency of tank 

farm data bank entries:2 

2

 The Environmental Policy Institute (EPI) obtained the data bank in about 1983 as a result of a Freedom of 

Information Act request. EPI no longer exists due to a merger, and the document is no longer available. It covered 

the period from late 1953 to 1982. SC&A requested it from NOISH as part of the SRS document request, but it was 

not available. A visit to SRS was conducted in February 2007 to provide for a firsthand review. 

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Table 4-5. Annual Average Number of Entries into the F- and H-Area Tank Farm 

Data Bank in Various Periods 

Period Average number 

of entries per year 

Comments 

(added in this review, not part of the table in the reference) 

1953–1959 4 

Spills and other incidents not recorded; no entries showing 

worker exposures in this period, though some incidents and 

conditions with high radiation rates are reported. 

1960–1965 32 First explicit worker radiation dose estimate entries are from this 

period. 

1966–1969 85 

1970–1976 290 

1977–1982 1,800 

Increase is mostly in items such as instrument maintenance and 

other entries not containing worker dose data. Many entries 

contain worker dose data. 

It is clear that there were substantial changes in the frequency of entries into the data bank. This 

does not necessarily indicate a corresponding increase in the frequency of incidents. Rather, it 

appears likely that more inclusive criteria for making entries into the data bank were adopted 

over the decades. Since many early incidents, including spills of high-level radioactive waste, 

were not recorded in the data bank, and since the Special Hazards Investigation (SHI) index is 

also incomplete, as acknowledged by SRS, it raises the question of how complete the record of 

incidents might be in terms of individual worker dose records, at least for tank farm workers. 

This is a crucial issue, since the NIOSH dose reconstruction procedure for SRS relies heavily on 

the essential completeness of DOE dose records and looks to the computer-assisted telephone 

interview (CATI) as a supplement. At least in the case of the F- and H-Area Tank Farms, this 

assumption needs to be verified. 

Two steps are necessary to verify the completeness of incident information provided to 

NIOSH/ORAUT for dose reconstruction. The record of known incidents in various data banks, 

worker records, SHI reports, and incident reports should be compared to the master incident list. 

Second, the master incident list needs to be scrutinized for completeness through review of 

records, interviews with site experts, and statistical analysis. This appears essential, since it is 

clear that outside workers, such as those in the high-level tank farm areas, repeatedly and 

frequently encountered conditions with high radiation rates of several roentgens per hour, tens of 

roentgens per hour, and sometimes even hundreds of roentgens per hour (Makhijani et al. 1986, 

Tables 1 through 11). 

Given the paucity of entries in the F- and H-Area Tank Farm data bank, the problem of 

inadequate or missing data regarding incidents may be especially acute in the early years. 

In this context, and pending further investigation, it would be reasonable to apply the term 

“early years” to mean the period from the inception of tank farm operations to at least 

1965 and probably to the end of the 1960s. An evaluation is needed as to whether the term 

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should be extended to the mid-1970s with regard to missing incidents. To SC&A’s 

knowledge, NIOSH has not made a master list of incidents available to dose reconstructors 

for reference. 

4.4.1.3 External Exposure Geometry Issues Related to Tank Farm Workers 

Employees working in the F- and H-Area Tank Farms experienced a large variety of exposure 

geometries. Determination of the location of the badge in relation to geometry of exposure is 

especially important in the tank farm area due to the highly nonuniform nature of the 

maintenance work done there, frequent high radiation rates, areas with spills, and other 

contamination having very site-specific contamination geometry. For instance, in repair and 

maintenance work performed on piping, in junction boxes, and on other tank farm equipment, as 

well as during cleanup after spills of high-level waste, multiple badges would be essential to a 

sound estimate of organ dose. 

Dose reconstruction for tank farm workers would therefore appear to face significant issues of 

technical accuracy and possibly data adequacy with regard to external dose because of the 

following factors, none of which are discussed in the TBD: 

•	 The location of the exposed organ relative to the source of radiation compared to the 

location of the badge(s) 

•	 Whether multiple badges were used 

•	 What entries were made into the records when multiple badges were used 

NIOSH proposed an exposure scenario analysis using the ATTILA code for a situation 

where the source of radiation originates underneath the workers, with selected 

examination of various geometry parameters and source terms. SC&A notes that the 

Mallinckrodt ATTILLA run for a similar situation showed that the correction factor was 

significant. The completion of this work should be expedited to reduce the reworking of 

claimant dose reconstruction cases. 

Although SRS had an established multiple badging program, it is unclear whether multiple 

badging was used during tank farm work. NIOSH/ORAUT should investigate the specific 

exposure conditions of the tank farm workers and evaluate the incident exposure versus the 

badge location. 

4.4.1.4 Radiological Zone Designation 

In addition to these issues, questions remain regarding how the various radiological control areas 

were designated and how such designations were changed over time. SC&A understands from 

site expert interviews that some parts of the F- and H-Area Tank Farms were designated as 

radiation zones, but that the entire F- and H-Area Tank Farm was not designated as such (SC&A 

2005a). Given that incidents may have been missed due to lack of recording, the potential for 

the significant exposure of workers who were in radiologically contaminated areas that were not 

designated as such needs to be investigated. The TBD does not discuss this issue. The 

importance of this and similar issues arises from the fact that the TBD assumes that unmonitored 

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workers were those unlikely to encounter radiation areas. The validity of this assumption needs 

to be checked against actual historical practices of contamination of outdoor areas, as well as 

changing definitions of radiological control areas over time. 

4.4.1.5 Comments on Completeness and Adequacy of Data Relating to F- and H-Area Tank 

Farm Workers 

The above discussion indicates that NIOSH had not evaluated site data, including the crucial 

tank farm data bank and the master list of incidents, when preparing the original TBD and 

Revision 03. Hence, NIOSH’s data evaluation is incomplete with regard to tank farm exposure 

conditions. 

SC&A’s evaluation of the tank farm data bank (based on summaries of entries in Makhijani et al. 

1986) indicates the waste management Fault Tree Data Bank is critical to addressing exposures 

associated with the tank farm operations. This compilation of data is unique to the DOE 

complex and provides valuable references for nonroutine exposures to SRS workers (Minnick 

and Wellmaker 1995). Unlike other sites where incident data are scattered in various sources, 

SRS has assembled a centralized databank that can aid claimants and NIOSH. Some 35,000 

incidents are documented from “60 types of sources, 47 Westinghouse Savannah River Company 

(WSRC) publications and 6 unpublished sources such as log books” (Minnick and Wellmaker 

1995, p. 4) spanning the period 1951 to 1994. The database includes “...incidents with 

significant potential for injury or contamination of personnel” (Baughman et al. 1993, p. 6). For 

instance, in 1993 a status report and description of this database contained search results for 

incidents of interest to NIOSH as reproduced in the following table: 

Table 4-6. Statistical Data from the tank farms Fault Tree Data Bank by Category. 

Number of Category Date Range Trend Best Fit Incidents 

Fires 79 1956–1992 Increasing Weibull 

Skin Contamination 142 1980–1992 Increasing Weibull 

Transfer Errors 73 1961–1992 No Change Weibull 

Overflows 101 1961–1992 Increasing Lognormal 

Uncontrolled Reactions 10 1977–1992 Increasing Weibull

 Source: Baughman et al. 1993, Table 2. 

The data bank has been used for several purposes at SRS, including radiation incident analyses. 

It has been used to determine the frequency and severity of radiation releases involving potential 

exposure to workers. In 1995, an analysis of tritium releases and fires was performed on 13,000 

incidents that took place at SRS 232-, 233-, and 234-H facilities from the 1950s to the early 

1990s. The study identified 37 fires at these facilities involving releases greater than 1,000 

curies of tritium (Wellmaker 1995). This data bank has been made available to SRS site 

contractors and well as to other groups, including Risk Assessment Corporation (RAC), which 

conducted radiation dose reconstruction for the Centers for Disease Control and Prevention 

(CDC). A user’s guide for the data bank also has been developed (WSRC 1995). Default dose 

reconstruction models for tank farm workers, as proposed by NIOSH/ORAUT in response to the 

SC&A finding, should not be developed in a vacuum that does not account for incident data 

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providing radiological exposure data. Although NIOSH/ORAUT’s position is that 

“...significant dose to workers from incidents are already contained in the dose of record 

information...” (Minnick and Wellmaker 1995), this does not necessarily mean that such records 

are accurate or complete. 

The Work Group, through NIOSH, made arrangements for an onsite review on February 28– 

March 1, 2007, by representatives from NIOSH, ORAUT, SC&A, and the Work Group of what 

was thought to be the tank farm Fault Tree Data Bank. The database reviewed is referred to as 

the WSMS incident database, which was developed by Bill Durant for safety analysis of 

activities in the separations area. In general, the information in the database appears to be 

primarily from the 200-F and 200-H Areas with minimal incident information from the reactor 

areas, heavy water area, the Savannah River Technical Center, the raw materials area, and other 

site operations. Over 90% of the entries were related to the separations area operations, 

including tritium operations. The discussion under Comment 13 provides a more detailed 

description of the database. 

The entries in the WSMS incident database were compared to those abstracted from the tank 

farms Fault Tree Data Bank documented in Makhijani et al. (1986). Most of the entries reviewed 

from the tank farms Fault Tree Data Bank were not located in the WSMS incident database. 

This may reflect a difference in operational scope between the two collections of incident data. 

SRS requested Central Records to search the records database for “Fault Tree.” However, the 

observations made to this point imply that multiple Fault Tree Data Banks have been maintained, 

corresponding to the various areas of the site. For example, the raw materials area has its own 

data bank. SRS did not originally provide the results of the search to NIOSH/ORAUT or SC&A, 

but NIOSH/ORAUT has recently requested this information. 

The tank farms database can serve to determine the assumptions that would be suitable for giving 

claimants who worked in the tank farms the benefit of the doubt in the face of considerable 

uncertainties. The lack of evaluation of primary data sources has left the TBD without a realistic 

way to estimate uncertainties. These problems are likely to be especially acute for the early 

years. The TBD radionuclide list is not complete for reasons that are not clear. The lack of 

clarity arises from the absence of any references or analysis in relation to the radionuclide lists 

that were chosen for the F- and H-Area Tank Farms. 

It is unclear whether adequate data are available to reconstruct any but the minimum doses for 

tank farm workers because of the various issues, data gaps, and uncertainties discussed above. 

The situation regarding early workers is especially unclear. A judgment on this question will be 

difficult or impossible without a careful evaluation of the available literature and an 

accompanying analysis of radiological conditions, exposure potential, and issues related to 

whether multiple badging was prevalent in tank farm work and, if so, how the data were 

generated and entered into dose records. 

These tank farm findings may also have implications for other areas of outdoor work, such as the 

burning ground and seepage basins. The TBD has no discussion of the former and no analysis 

relating to dose reconstruction of the latter. 

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4.4.2 Applicability to Revision 03 

Comments related to the tank farms are still applicable to Revision 03. 

4.4.3 Summary of Overall NIOSH Response 

Internal Dose 

For internal dose estimation, the radionuclides of concern were identified on the basis of those 

that would deliver the majority of the dose and those that could be used as tracers. Both the 

annual and the committed effective dose equivalents were used. The source of the list of 

radionuclides of concern for the tank farms, as well as most of the radionuclide lists for the other 

facilities, is found in Facility Descriptions (LaBone 1996), as is mentioned in Section 1.3 of the 

site profile following the list of tables. During the period of active reprocessing of fuel, the tanks 

would have contained many different fission products. 

The significance of the radionuclides listed in Facilities Descriptions was determined as follows. 

“The radionuclide contents of the waste tanks shown in Table 17.3 through 17.6 [references are 

to tables in Facilities Descriptions] were reviewed and the content of the combined tanks 

(Table 17.3) was chosen as being a representative sample for design purposes. The radionuclides 

of concern were identified on the basis of those that would deliver the majority of the dose and 

those that could be used as tracers.” 

Table A-14 of the TBD lists Cs-137 and Ru-106 as significant in terms of total activity but these 

radionuclides do not produce as much dose as Sr-90, Ce-144, and Cm-244. However, 

NIOSH/ORAUT indicates that Cs-137 and Ru-106 can easily be mentioned in the brief 

description in Chapter 1 of the TBD. NIOSH will add Ru-106 to the list of H-Area radionuclides 

important to external dose, and remove Pu-238. U-233, Tc-99, and Th-232 are not listed in 

Table A-14 of Revision 03 of the TBD as significant in terms of either total activity or dose, nor 

would they be expected to be significant relative to other high-level waste products or actinide 

wastes. Zr-95/Nb-95 are quite abundant in freshly irradiated fuel but tend to be less abundant in 

the waste tanks because of the shorter half-lives relative to Ce-144, Ru-106, Cs-137, and others. 

They also have smaller organ dose conversion factors. 

Attachment A from the June 5, 2006, matrix provides a response regarding the inadequacy of 

incident records for tank farms in the 1950s and 1960s. 

External Dose 

Section 5.6 of Revision 03 of the TBD, “Organ Dose,” is a summary section of recommendations 

to the dose reconstructor to be used in dose reconstruction. Since shallow dose is not addressed 

in this Site Profile, tank farm operations do not present unique sources of radiation exposure that 

are not already identified in the existing Site Profile. 

The dose of record information already documents significant doses to workers from incidents in 

the dose of record information. In addition, Form OCAS-INT-004 requests DOE to submit 

relevant incident information for consideration in dose reconstruction. 

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4.4.4	 Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Action Items: 

•	 Provide the Board with an expanded review of the adequacy of its SRS radionuclide list 

by referencing additional source term information [e.g., Fault Tree Data Bank 

compilation (NIOSH to obtain from DOE)]. 

•	 Provide the Board with its specific comments related to issues raised regarding 

Attachment A, as identified in SC&A comments (and elaborated on in a separate SC&A 

note). This review should address gaps in tank farm radionuclide characterization and 

incidents (including scenarios). 

•	 Provide the Board with its evaluation of exposure geometry in tank farm work, 

encompassing common tasks such as cleaning up spills and work on pipes and valves 

where the source was at a lower level than the badge. 

•	 Pull data (e.g., from the tank farm Fault Tree Data Bank registry) and verify 

representativeness and claimant favorability. 

4.4.5	 Closure Status 

The adequacy of the F- and H-Area Tank Farm characterization in the TBD is questionable for 

use as dose reconstruction guidance. Data evaluation appears to be incomplete with regard to 

exposure conditions, radionuclides of concern, and uncertainty. This is particularly true for early 

periods of operation, where primary records involving key operations and incidents are lacking. 

The tank farms database, not currently evaluated by the TBD, can serve to determine the 

assumptions that would be suitable for giving claimants who worked in the tank farms the 

benefit of the doubt in the face of considerable uncertainties. The lack of evaluation of primary 

data sources has left the TBD without a realistic way to estimate uncertainties. The potential for 

internal and external exposure for unmonitored workers in areas not designated as radiological 

control areas needs to be investigated. Revision 03 of the site profile included no changes in the 

discussion regarding the tank farms. Proposed Revision 04-E (Scalsky 2006) included only 

minimal changes, such as the addition of Cs-137 and Ru-106 to Table A-14. Default intakes for 

tank farm workers for different periods of time were added, including those for actinides. This 

revision does not resolve all issues associated with the tank farms. It is recommended that this 

issue remains open pending completion of Work Group action items and the release of proposed 

Revision 04-E (Scalsky 2006). 

4.5	 COMMENT 5: LACK OF COMPREHENSIVE EVALUATION OF EARLY 

MONITORING PROGRAM 

For early SRS workers, the site profile lacks a comprehensive evaluation of the early monitoring 

program with respect to its consistent application, adherence to procedures, recordkeeping, and 

adequacy, all of which hold significant implications for reconstructing doses for unmonitored 

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workers during the early years. Although early procedures clearly define when beta/gamma 

dosimeters were to be worn, the requirements for the use of neutron dosimeters and bioassay 

collection were left up to the field support organization. Gaps in data availability were noted for 

individual neutron exposure data, tritium exposure data, internal and external exposure data from 

special campaigns, and the radionuclide source term lists (and attendant concentrations and 

activity levels) used in the TBD, including those for the tank farms, RU, and environmental 

releases. As the basic approach for dose assessment for early workers, NIOSH proposed to use a 

maximizing dose for noncompensable claims and coworker dose for unmonitored, potentially 

exposed workers. Assignment of missed neutron dose is based on the application to the photon 

dose of a neutron-to-photon ratio prior to 1971. 

4.5.1 Issue Description 

The Radiological Control Organization was not centralized at SRS until more recent times. 

DPSOP-40, Operating Procedure for Radiation and Contamination Control, outlined the basic 

requirements that were to be followed with respect to personnel monitoring. In essence, field 

support personnel determined the requirements for routine and special bioassay and dosimetry 

using the following guidelines, as set forth in DPSOP-40 (DPSOP 1959 and 1960): 

Film badge dosimeters are to be worn at all times by all personnel in exclusion 

areas, Regulated Zones, or Radiation Danger Zones (RDZs). 

Pocket meters are to be worn by all personnel where exposure rate is 25 mr/hour 

or greater, or when specified on the Special Work Permit. 

Neutron film badges or TLNDs are worn when specified by Health Physics on 

jobs where personnel are exposed to neutron radiation. 

Neutron pencils are worn when specified by Health Physics on jobs where 

personnel are exposed to slow neutron radiation. 

All personnel working in Regulated Zones or Radiation Danger Zones are 

periodically checked for assimilation of radioactive material. In buildings in 

which tritium is present, bioassay samples are submitted as directed by Health 

Physics. 

Special bioassay samples may be requested by Health Physics through the 

employees’ supervision, when a suspected assimilation of radioactive material 

occurs. 

Prior to 1959, the Operating and Health Physics departments had to approve bioassay and 

dosimeter requirements (DPSOP 1953 and 1956). Work permits and facility-specific procedures 

were used to supplement the requirements of DPSOP-40. These requirements were documented 

on a Special Work Permit for nonroutine jobs. In 1971, requirements for in vivo and in vitro 

bioassay were outlined in DPSOL-193, Health Protection Procedures, or by specific request 

from Health Physics (DPSOL 1971). The requirements for external monitoring were the same as 

defined above (DPSOP 1971, 1974, and 1976). By the late 1980s, the dosimeter and bioassay 

requirements were clearly outlined in DPSOL-193. A beta/gamma dosimeter was required for 

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all personnel who handled radioactive materials or entered facilities where radioactive material 

was handled or stored. TLNDs were required when the neutron dose rate was equal to or greater 

than 1 millirem (mrem)/hour (DPSOL 1989b). Routine and special bioassay requirements were 

also outlined for both in vitro and in vivo counting (DPSOL 1987, 1988, and 1989a). As is noted 

in the monitoring requirements listed above, facility personnel and not a central organization 

initially determined neutron, tritium, and special monitoring. This raises questions of 

consistency in monitoring for the early years. 

In February 1999, SRS underwent an independent assessment of its internal dosimetry program. 

One of the findings is stated, as follows (WSRC 1999): 

Facility personnel did not consistently adhere to the WSRC procedural 

requirements for initiating special bioassay sampling. In addition there was no 

mechanism in place for ensuring subcontractors submitted termination bioassay 

samples. 

One of the corrective actions implemented by SRS as a result of the audit was to make the 

Internal Dosimetry Group responsible for determining bioassay requirements. 

Without a single organization determining requirements for neutron and bioassay requirements, 

there may have been inconsistencies from area to area in who was monitored and who was not 

monitored. Some consideration should be given to the following: 

•	 Early monthly progress reports from the Works Technical Department include statistics 

by area of who was monitored for tritium. Initially, according to the reports, all the 

tritium monitoring was listed under F-Area, with no monitoring indicated for the reactor 

areas. Starting in 1957, monitoring was conducted for F-Area and H-Area workers. It 

was noted that no tritium analysis was documented in these reports for reactor workers 

during the 1956–1960 timeframe (DuPont 1957, 1958, 1959, 1960, and 1961). Based on 

SRS dose reconstructions reviewed by SC&A, it is evident that bioassay samples were 

collected from some reactor workers in this time period. The adequacy of the tritium 

monitoring in the reactor areas and heavy water rework area should be evaluated and the 

potential for exposures compared to the missed dose assignments outlined in ORAUTOTIB-0011, Technical Information Bulletin: Tritium Calculated and Missed Dose 

Estimates (Siebert 2004). 

•	 Review of early handwritten monitoring records (1954–1957) indicated that neutron 

monitoring was intermittent rather than routine in some areas. 

•	 The progress reports indicate that uranium bioassay was initiated in 1953, plutonium 

bioassay was initiated in 1954, and tritium and fission product bioassay were initiated in 

1955. Some thorium bioassay appears to begin in 1956. In 1960, mention of “Special 

Product” bioassay begins to appear in the reports. Americium-curium statistics on a 

number of monitored workers appear starting in early 1964. This information should be 

reviewed thoroughly with respect to identifying those exposed versus those monitored in 

these earlier periods of time. 

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Furthermore, monthly reports from the Works Technical Department contain information 

indicating that the tritium-monitoring program may not have been comprehensive. These reports 

track the number of tritium urinalysis samples by area. ORAUT-OTIB-0011 (Siebert 2004) 

provides a process for calculating missed dose based on the detection limits in bioassay 

sampling, but this may not be bounding for unmonitored individuals exposed to tritium. 

Although early procedures clearly defined when beta/gamma dosimeters were to be worn, the 

requirements for neutron dosimetry were determined by the field support organization and may 

not have been consistent across the plant site. Figure 5.3.4.2-2 of the SRS TBD shows the trends 

in SRS and Hanford collective neutron dose normalized to plutonium production. The table 

clearly shows that there was minimal collective neutron dose prior to 1964 (Scalsky 2005). 

Review of the monthly reports from the Works Technical Department for December 1956, 1957, 

1958, 1959, and 1960 indicates that 1,776, 1,001, 1,805, 2,023, and 2,778 NTA badges were 

processed for each year, respectively (DuPont 1957, 1958, 1959, 1960, and 1961). During this 

period of time, dosimeters were exchanged on a weekly schedule. The number of badges 

processed likely included multiple badges for an individual. From this information, it appears 

that minimal numbers of individuals at SRS were monitored for neutron exposure. A majority of 

these individuals were likely from the FB-Line and HB-Line where the plutonium finishing 

process occurred. NIOSH should investigate the completeness of the neutron monitoring at each 

facility for which there was a neutron potential. Such an investigation would be helpful to dose 

reconstructors in determining when an individual may have experienced neutron exposures but 

were not monitored. 

For noncompensable claims, Revision 04-E (Scalsky 2006) of the site profile recommends 

applying the 95th percentile neutron-to-photon dose ratio to the recorded and missed photon dose 

for workers routinely present in work areas prior to 1972 with a potential for neutron exposure. 

For likely compensable claims, the geometric mean value of the neutron-to-photon dose ratio is 

applied. The four main areas at SRS with potential for neutron exposure include the plutonium 

facilities in the 200 Area; the Calibration Facility (736-A) and the Cf-252 Facility (773-A) in the 

700 Area; reactors in the 100 Area; and Building 321 (plutonium-aluminum alloys) in the 300 

Area. 

It is not clear whether missed neutron dose is assigned to all job classifications, or how NIOSH 

determined whether an individual worked in an area with potential neutron exposure. This is 

especially concerning for the mobile segment of the work force. Any reliance on previous 

neutron monitoring as an indicator of potential exposure would require the availability of all 

neutron data. 

Database Gaps 

ORAUT has developed ORAUT-OTIB-0032, External Coworker Dosimetry Data for the 

Savannah River Site (Merwin 2006) for the assignment of external dose to workers without 

monitoring data but with the potential for exposure. This would include workers who would 

have been monitored under current standards but were not, or workers who have incomplete 

dosimetry data. These external beta/gamma coworker doses are proposed for use in cases were 

early worker monitoring data are incomplete. NIOSH used dosimetry data for monitored SRS 

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workers from the HPAREH database. The annual data in HPAREH were prorated for partial 

years of exposure. 

External radiation exposure data are available in three electronic files commonly referred to as 

the Fayerweather file, HPAREH and SRPABST. The three files do not contain radiation 

exposure information for all individuals employed at SRS. In fact, the Fayerweather and 

SRPABST files were compiled for particular studies conducted by DuPont and ORAU, 

respectively, and include only individuals from specific cohorts. HPAREH, described below, 

contains only those workers who terminated in 1979 and after (Richardson et al. 2006, 

Robertson-DeMers 2003). For all remaining individuals, dose information has to be obtained 

from the hardcopy records (e.g., logbooks). As SRS researches individuals terminating prior to 

1979, they add this exposure information to the HPAREH file. This is why HPAREH contains 

data regarding some individuals terminating prior to 1979, although it does not include a 

majority of the monitored population terminating prior to 1979. The HPAREH and successive 

electronic dosimetry files are the only sanctioned radiation exposure files at SRS. The hardcopy 

records contain the most complete set of dosimetry data available. For the cohort of 18,883, 

15,752 annual dose records were identified from a review of dosimetry logbooks, which were not 

in a computerized data system (Richardson et al. 2006). 

NIOSH is currently using only the HPAREH file to evaluate external coworker doses. HPAREH 

consolidated data from the personnel radiation exposure files, logbooks, and magnetic tapes. It 

was designed to provide annual radiation exposure reports to plant workers and DOE. There was 

no requirement to go back and reconstruct doses for terminated workers. As a result, SRS 

focused their computerization of data on individuals who were actively working at the site in 

1979 and thereafter. For those individuals included in HPAREH, exposure for all years of 

employment is available. In some cases, a cumulative dose is recorded in the first year of data in 

HPAREH, rather than the annual dose by year up to that point. Each year, the Health Physics 

Master System is used to update the HPAREH database with additional exposure information, 

terminations, and other information (Taylor et al. 1995). As individuals or other entities 

requested exposure records, SRS compiled the individual’s dose history and entered the 

information into HPAREH. During the implementation of HPAREH, SRS made a concerted 

effort to resolve radiation exposure history gaps and inconsistencies. 

ORAUT-OTIB-0032 (Merwin 2006) uses a version of the HPAREH file as a basis for 

determining coworker dose by year. NIOSH described the validation of the data selected for 

coworker dose development as follows (Merwin 2006): 

The validity of the data selected for coworker dose development was confirmed by 

selecting a sampling of HPAREH summary data submitted by the site as a part of 

the EEOICPA Subtitle B program and comparing it to the data described above. 

A review of the data for two claimants with complex and extensive dosimetry 

records covering the years 1953 to 1999 indicated good agreement between the 

two data sets. Importantly, when the data did not match exactly, there was no 

apparent bias toward an under or overestimate of dose for either data set, 

suggesting that relying on the HPAREH annual data for coworker dose 

reconstruction would not result in a negative bias against the claimants. 

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However, SC&A contends that the review of hardcopy results verses HPAREH results for two 

individuals is not an adequate sampling for validation of the database. The systematic gaps in 

HPAREH occur for individuals terminating prior to 1979. The two individuals worked at SRS 

through 1999 and would not fit into this category of workers. 

NIOSH needs to evaluate whether using HPAREH exclusively for determining coworker dose is 

appropriate. The basis for considering the HPAREH data to be sufficient for the assignment of 

coworker dose prior to 1979 requires further explanation since this file includes only a portion of 

the population. NIOSH should demonstrate that data for the assignment of coworker dose are 

representative of all time periods and work groups. 

4.5.2 Applicability to Revision 03 

NIOSH/ORAUT has established methods to assign dose to unmonitored workers in Revision 03 

of the site profile and the following supplemental documents, which were reviewed in 

preparation of this report: 

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

(November 7, 2006)—Provides information to allow dose reconstructors to assign doses 

to SRS workers who have no or limited monitoring data, based on site coworker data. 

(Merwin 2006) 

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

(July 15, 2003)—Describes the maximum internal dose estimate for SRS claims. 

(Brackett 2003) 

•	 ORAUT-OTIB-0011, Technical Information Bulletin: Tritium Calculated and Missed 

Dose Estimates (June 29, 2004)—Documents the method for estimating tritium missed 

and calculated doses from urine. (Siebert 2004) 

4.5.3 Summary of Overall NIOSH Response 

NIOSH has established methods to provide doses to unmonitored workers, and new guidance is 

being developed for unmonitored internal dose that clarifies intake radionuclides and levels at 

the various facilities. 

4.5.4 Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Provide the Board with a description of classes of workers who were not appropriate 

monitored or time periods when workers were not appropriately monitored during the 

early phases (e.g., no early H-3 bioassay in reactors, early exposure gaps in HPARAH 

database, tank farms). 

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SC&A Actions: 

•	 Provide an inventory of data sources used in the preparation of the individual dosimetry 

data submittals. This inventory will assist in questions related to the monitoring of early 

workers. 

4.5.5	 Closure Status 

An evaluation of the comprehensiveness of the early monitoring program should be completed 

for early workers to determine whether existing site profile methodologies bound their dose. 

This is especially important in the case of workers who were not monitored but were exposed to 

a radiological hazard. Without a single organization determining neutron dosimeter and bioassay 

requirements, actual practices may have been inconsistent. The inconsistencies may also include 

the request for special interpretations of film badges for workers in the plutonium areas. The 

adequacy of the early monitoring program (i.e., who received monitoring) will not be resolved 

exclusively by an inventory of records provided in claimant files. References such as incident 

databases, monthly reports, and dosimetry logs also need to be considered. Additional validation 

of the HPAREH database as the exclusive source for external coworker dose determination, 

given the incompleteness of early data, is necessary to demonstrate this data is adequate for this 

use. It is recommended that this finding be considered open pending completion of Work Group 

actions provided to NIOSH. The discussion for Comment 14 includes additional information on 

data sources available for early workers, and Comment 9 discusses the application of a 

maximizing internal dose to unmonitored workers. 

4.6	 COMMENT 6: HIGH-FIVE APPROACH INCONSISTENT WITH THE 

METHODOLOGIES RECOMMENDED IN 42 CFR PART 82 FOR 

CALCULATION OF INTERNAL DOSE 

The high-five approach, as currently explained in ORAUT-OTIB-0001 (Brackett 2003), is not 

consistent with the methodologies recommended in 42 CFR Part 82 for the calculation of internal 

dose. The dose reconstruction process must comply with the requirements of 42 CFR Part 82. 

To provide a method for effectively implementing these requirements, NIOSH has written 

technical guidance documents on external and internal dosimetry. While ORAU has committed 

to the use of these guidance documents in its quality assurance program plan, the use of the highfive approach to assign internal dose is not consistent with this guidance. 

The method outlined in ORAUT-OTIB-0001 (Brackett 2003) to assign internal doses is based 

upon a hypothetical intake with the following characteristics (Brackett 2003, p. 3): 

All radionuclides for which internal deposition by inhalation was calculated by 

the Savannah River Site were reviewed, except for tritium, which is addressed 

separately. 

The amount of the inhalation intake for each radionuclide is the average (mean) 

of the five largest documented intakes, or the average of all intakes if there were 

fewer than five intakes reported for a radionuclide. 

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An acute inhalation intake was assumed to have occurred on January 1 in the first 

year of employment. 

ICRP 66 and 68 modeling and default parameter values were used to determine 

dose. 

The material type resulting in the largest dose to the organ or tissue of interest 

was used. This was typically the most soluble form of the material because it 

would clear from the lung more rapidly than insoluble material, thus depositing 

in the organ or tissue sooner. 

...Intakes and doses at SRS were calculated using regulatory-prescribed ICRP 30 

methodologies rather than the newer ICRP methodology prescribed for this dose 

reconstruction effort. The material classes used in the calculations were based on 

workplace source term information or the class that provided the best fit to the 

bioassay data; the most claimant favorable class was not necessarily selected. 

As clearly stated above, SRS uses ICRP 30, Limits for the Intake of Radionuclides by Workers 

(ICRP 1979), to calculate the intakes used in the high-five approach. The organ dose is then 

calculated using modeling and default values from ICRP 66, Human Respiratory Tract Model for 

Radiological Protection (ICRP 1994), and ICRP 68, Dose Coefficients for Intakes of 

Radionuclides by Workers (ICRP 1995). The regulations in 42 CFR 82.18(b) direct NIOSH to 

do the following: 

...calculate the dose to the organ or tissue using the appropriate current 

metabolic models published by the ICRP. 

Furthermore, in the question-and-answer section accompanying 42 CFR Part 82, NIOSH 

discusses the use of ICRP models: 

As explained in the interim final rule and above, NIOSH is using current ICRP 

models because they represent improvements in the science of internal and 

external radiation dosimetry compared to older ICRP models. 

The intake quantity is the basis for determining final organ or tissue dose. In the case of SRS, 

NIOSH has decided to use intake information calculated using ICRP 30 methodology. Since the 

issuance of ICRP 30, ICRP has developed the new lung model outlined in ICRP 66 and the 

revised dose coefficients in ICRP 68. Therefore, ICRP 30 is not the most current metabolic 

model and is not consistent with the direction provided by 42 CFR 82.18(b). 

4.6.1 Applicability to Revision 03 

This issue is applicable to Revision 03 and ORAUT-OTIB-0001 (Brackett 2003). 

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4.6.2	 Summary of Overall NIOSH Response 

NIOSH has indicated that it expects to complete the process of completing the update to the 

high-five approach in December 2006. Information in the revised TBD refers the dose 

reconstructor to ORAUT-OTIB-0001 (Brackett 2003), which describes the high-five approach. 

The updated approach along with any modifications to the TIB will be reviewed by SC&A when 

provided by NIOSH. 

4.6.3	 Work Group Action 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

• Complete the update of high-five intakes using new models and provide this to the Board. 

4.6.4	 Closure Status 

As noted above, information in the revised TBD refers the dose reconstructor to ORAUT-OTIB0001 (Brackett 2003), which describes the high-five approach. NIOSH indicated that it would 

update the high-five approach and base revised calculations on bioassay data rather than data in 

the SRS Internal Dosimetry Record. This update was expected in December 2006 but has not 

been completed as of August 2007. Until ORAUT-OTIB-0001 (Brackett 2003) is revised and 

the TBD references the correct TIB, regulatory issues will continue to be associated with the 

application of the high-five approach. Using urinalysis as a basis for dose calculation will 

eliminate the use of intake data derived with ICRP 30 and allow for the exclusive use of ICRP 60 

methodology or more current methods. Revision 03 (Scalsky 2005) and Revision 04-E (Scalsky 

2006) of the TBD still reference the 2003 procedure on maximizing internal dose. The updated 

approach along with any modifications to the TIB will be reviewed when provided by NIOSH. 

It is recommended that this issue remains open pending the release of the revision to ORAUTOTIB-0001 (Brackett 2003) and subsequent reflection in Revision 04-E (Scalsky 2006) of the 

TBD. 

4.7	 COMMENT 7: LIMITATIONS ASSOCIATED WITH THE ASSIGNMENT OF 

OCCUPATIONAL ENVIRONMENTAL DOSE 

The method used to reconstruct doses to unmonitored outdoor workers due to airborne emissions 

employs an atmospheric dispersion model, assumptions, and a resuspension factor that do not 

appear to be claimant favorable and are not entirely appropriate for this class of problem. For 

modeling of airborne radionuclide releases, one potentially significant issue is the 

nonconservatism of the standard Gaussian model used in the TBD where it pertains to 

“nonstandardized” short-term releases occurring during stable atmospheric conditions. Based on 

an SC&A review of the literature, it also appears that the TBD resuspension factor of 1×10-9 per 

meter may not be claimant favorable. 

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4.7.1 Issue Description 

The fundamental approach employed in the environmental dose section (Chapter 3) of the site 

profile for deriving occupational environmental doses uses (1) a sector-averaged Gaussian plume 

model, (2) source terms and a radionuclide list originally intended for offsite dose estimation, 

and (3) a resuspension factor of 1×10-9 per meter for estimating air dust loading due to 

radionuclides in the soil. For the purpose of deriving outdoor doses to unmonitored workers 

from airborne emissions, NIOSH employed the source terms reported in the summary report 

prepared by Cummins et al. (1991) and the dose reconstruction report prepared by RAC, RAC 

Report No. 1-CDC-SRS-1999, Savannah River Site Dose Reconstruction Project Phase II: 

Source Term Calculation and Ingestion Pathway Data Retrieval, Evaluation of Materials 

Released from the Savannah River Site (CDC 2001). The atmospheric source terms reported in 

these documents appear to be limited to monitored releases, and they are reported in terms of 

total annual releases by year for the purpose of deriving historical offsite doses. This strategy 

may not be entirely applicable to reconstructing the doses to onsite workers for a number of 

reasons. Unmonitored and episodic releases that occur over a relative short period of time (e.g., 

days) may deliver relatively high doses to nearby outdoor workers that may have been missed. 

In addition, the application of average annual atmospheric dispersion factors based on standard 

Gaussian models may not be appropriate for exposures occurring close to the source term, 

especially to ground-level and/or episodic releases. 

For example, Chapter 4 of the RAC report (CDC 2001) refers to a report by Miller (1956), which 

presents information on releases due to incidents and accidents. The RAC report explains that 

these releases appeared to have been captured in the annual estimates of the source terms. 

However, it may be instructive to review these incidents from the perspective of the potential 

doses to onsite workers. The site profile would benefit from a more in-depth analysis of these 

issues, or at least a demonstration that the doses to workers from episodic and ground-level 

releases could not have contributed significantly to the doses to onsite workers, as compared to 

the doses derived in Chapter 3 of the site profile. 

SC&A has also noted that NIOSH/ORAUT has not made comprehensive use of information 

available relating to environmental releases at SRS. SRS published a series of reports providing 

a summary of releases from SRS facilities, including atmospheric and liquid releases, transport 

mechanisms, and concentration on and in the vicinity of SRS. These environmental reports 

include information on releases of activation products, americium, cesium, curium, fission 

products, neptunium, noble gases, plutonium, radiocarbon, radioiodine, strontium, technecium, 

tritium, and uranium. 

The assessment of environmental dose did not mention a number of radionuclides that are known 

to have been released from SRS facilities. These radionuclides included Am-241/243, Br-82, C14, Ce-141/144, Cm-242/244, Co-60, Cr-51, Cs-137, Eu-154/155, I-133, I-135, Kr-85/85m, Kr87, Kr-88, Nb-95, Np-239, P-32, Ru-103, Ru-106, S-35, Sr-89/90, technecium, Th-232, Xe131m, Xe-133, Xe-135, Y-91, Zn-65, and Zr-95 (Carlton et al. 1995; Jannik 1997; WSRC 1996 

and 1997; Carlton et al. 1992a, 1996, 1993a, 1992b, and 1993b; Kantelo et al. 1993). Although 

the TBD indicates that the radionuclides selected for evaluation deliver 95% of the potential 

missed dose, it does not discuss the methodology used to determine which radionuclides and 

source pathways are important to onsite dose assessment. Screening calculations used for 

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developing an offsite radionuclide list may not be appropriate for determining an onsite 

radionuclide list. Some discussion is needed regarding the methodology used to determine 

significant radionuclides and pathways that are included in the determination of environmental 

dose. 

4.7.1.1 Dispersion model 

Chapter 3 of the site profile makes extensive use of the RAC report (CDC 2001) for deriving 

occupational environmental doses. In so doing, NIOSH has adopted the sector-average Gaussian 

plume model (using site-specific meteorological joint frequency data) to derive onsite exposures 

from both elevated and ground-level releases. This approach to reconstructing historical offsite 

doses, as performed by RAC in support of CDC, is generally scientifically sound for offsite dose 

reconstructions at sites that have generally flat or rolling terrain, and where the releases are 

relatively uniform over time. In addition, this approach can also be used for deriving offsite 

doses from episodic releases if the episodic releases were numerous during a given year and 

random over time. However, as discussed in the following paragraphs, this approach may not be 

appropriate for reconstructing onsite doses to workers for a number of reasons. 

SC&A recognizes that, within the framework of the approach it chose, NIOSH used some 

conservative assumptions for deriving occupational doses. For instance, NIOSH selected the 

highest sector-average annual atmospheric dispersion factor, which assumes that the worker is 

located year round downwind in the most prevalent wind direction at the site. However, certain 

fundamental issues associated with this approach could result in a substantial underestimate of 

the dose. Specifically, the Gaussian model breaks down in the near field for ground-level 

releases (i.e., those emissions that are released at a height that is less than about 2.5 times the 

height of the adjacent buildings). Under these circumstances, building wake effects cause 

turbulence that cannot be easily modeled by Gaussian methods. 

Another concern is that some workers may have been located downwind at the time of episodic 

ground-level or close to ground-level releases at a time when the meteorology may have been 

highly stable (i.e., very little dispersion). The estimated annual intakes in Attachment C of the 

TBD are based on average annual atmospheric dispersion factors, and the actual instantaneous 

meteorology could vary from these averages by several orders of magnitude on any given day. 

Under these circumstances, it may be more appropriate to employ the upper 95th percentile 

atmospheric dispersion factors for deriving doses, as opposed to the average annual atmospheric 

dispersion factors. This is the approach recommended for use by the U.S. Nuclear Regulatory 

Commission (NRC) for deriving doses associated with accidental releases from commercial 

nuclear power plants (NRC 1974). 

If large short-term releases occurred during stable conditions, such as during low windspeeds and 

stable atmospheric stability conditions (e.g., stability class E or F), the approach employed in the 

TBD could result in substantial underestimates of the doses to outdoor workers downwind from 

releases. The Hanford TBD acknowledges and explicitly addresses this issue and uses the 

RATCHET atmospheric dispersion code instead of the standard Gaussian model employed in the 

SRS site profile. SC&A suggests that the TBD revisit this issue and confirm that doses from 

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episodic releases were not, in fact, significantly underestimated because of the use of 

conventional Gaussian models. 

NIOSH’s response provided on June 5, 2006, did not resolve the short-term episodic exposure 

issues raised in that it did not address ground-level plumes from the open-pan burning of large 

amounts of contaminated reprocessing solvents. Based on a review of DuPont Savannah River 

Plant Procedure (DPSP) documents spanning 1956 to 1962, large amounts of spent solvent from 

the F- and H-Canyons were transported and stored in underground tanks. Tens of thousands of 

gallons containing significant concentrations of fission products and transuranics were stored and 

then piped to an open-pan incinerator, which burned approximately 1,000 gallons at a time. Panburning of spent reprocessing solvent appeared to occur every few months in the 1950s and 

1960s. In addition, several episodes involved environmental spills and leaks resulting in highbody dose rates as well as stack releases resulting in contamination of personal vehicles and 

large occupied areas on site. Modeling should address these episodic releases, particularly with 

respect to large particles. Offsite dispersion models used in the TBD (Scalsky 2006) are not 

viable for particle sizes above 0.5 micron and do not address the onsite deposition trajectory for 

large particles. Finally, ambient gamma dose rates to employees working outside and near the 

chemical separations operations were high in the 1950s and 1960s; and it is not clear if exposed 

workers wore badges. 

4.7.1.2 Annual Intake from Resuspension 

The methods available for deriving inhalation exposure from resuspended radionuclides include 

the resuspension-factor approach and the dust-loading approach. The dust-loading approach is 

used for those scenarios where information is available on the radionuclide concentration in 

surface soil dust (e.g., picocuries per gram (pCi/g)) and the airborne dust loading (g/m3

) of 

respirable-size particles. Using this approach, the product of the radionuclide concentration in 

the surface soil (pCi/g) with the dust loading of respirable-size particles (g/m3

) yields the 

airborne radionuclide concentration (pCi/m3

). This may be a suitable approach when dustloading data are available, because the radionuclide concentrations in soil are reported in terms 

of pCi/g. 

The resuspension-factor approach is used when information regarding the scenario is limited to 

surface contamination levels (e.g., pCi/m2

). Resuspension factors are empirically determined 

values expressed in units of pCi/m3 per pCi/m2 (which reduces to units of 1/m) for a given 

exposure setting. The product of the surface contamination level with the resuspension factor 

yields the equilibrium airborne radionuclide concentration (i.e., pCi/m3

). This is the approach 

employed in the site profile. 

NIOSH considers the resuspension factor of 1×10-9, chosen by Till (1983, p. 5-32), as the value 

most representative of undisturbed field conditions at SRS. The TBD does not define the basis 

for this assumption. The resuspension factor of 1×10-9, derived for offsite dose reconstruction, 

however, may not be representative of workers on site. These workers would be in closer 

proximity to contaminated soils and liquids than members of the public would be. In some 

cases, active work may be occurring at these sites that will generate more airborne dust and 

suspended particles. Consideration should be given to the determination of an appropriate 

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resuspension factor differentiating between a member of the public and an onsite, unmonitored 

worker. NIOSH should demonstrate that 1×10-9 is a claimant-favorable resuspension factor. 

Measured resuspension factors vary over very wide ranges. Kennedy and Strenge (1992) 

reported resuspension factors from approximately 1×10-11 to 1×10-2 m-1, which suggests that 

resuspension is a complex process of several parameters and that the specific conditions present 

at the time of measurement are critical. For modeling purposes, a resuspension factor is a 

lumped parameter that is used to account for a complex combination of mechanisms that are not 

very well understood but whose net effect is observed in the real world. 

The representative indoor resuspension data compiled by Beyeler et al. (1999) from experimental 

data and available information for indoor resuspension factors ranges from 2×10-8 m-1 (Jones and 

Pond 1964) to 1×10-3 m

-1 (Healy 1971). The reported data are generally from experiments that 

examined the resuspension of liquid or powder contaminated material that had been uniformly 

applied to clean surfaces in a laboratory-like setting. The highest values are typically associated 

with inefficient ventilation, excessive mechanical disturbance, or dusty conditions. Typically, 

the purpose of these studies was to help determine radiation protection safety guidelines for 

loose, residual, surface radioactivity. The original SRS TBD review (SC&A 2005a) provides a 

more detailed analysis of Beyeler’s summary. Indoor resuspension factor applicability to 

outdoor resuspension factors is questionable, but intuitively, one might expect outdoor 

resuspension factors to be generally higher due to wind and anthropomorphic activities that are 

likely to be greater outdoors than indoors. 

Several reports by Sehmel (1977 and 1980) revealed that there is enormous uncertainty in 

outdoor resuspension factors, as there is for indoor resuspension factors. In an investigation of 

resuspension factors at the Hanford reservation, Sehmel (1977) found outdoor resuspension 

factors ranging from 1×10-11 to 1×10-5 per meter. In a review of the literature on outdoor 

resuspension factors, Sehmel (1980) cites experimental studies where the values ranged from 

9×10-11 to 3×10-4 per meter for wind resuspension, and 1×10-10 to 4×10-2 per meter for 

mechanical stresses from human activities. He explains that there are many reasons for this 

variability, many of which relate to sampling and experimental techniques and the depth and 

nature of the contamination. For these reasons, the dust-loading approach is probably preferable 

when it can be employed. Stewart (1964) concludes that a representative value for a 

resuspension factor outdoors under quiescent conditions is 1×10-6 m

-1 and under conditions of 

moderate activity is about 10 times greater. 

Based on this review, it would seem that a resuspension factor of 1×10-9 per meter, as used in the 

TBD, may not be claimant favorable. An average value closer to 1×10-5 to 1×10-6 per meter 

would seem more appropriate for use in worker dose reconstruction, resulting in worker 

inhalation doses from resuspension that are 3 to 4 orders of magnitude greater than those derived 

in the site profile. 

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4.7.1.3 Dust Loading 

A report prepared for the NRC by Battelle Pacific Northwest Labs (Sutter 1982) provides a 

review of this subject. Table 4.7 summarizes some of the relevant information contained in that 

report. 

Table 4-7. Some Airborne Particulate Mass Concentrations 

Aerosol Mass Concentration Reference 

Dust storm 0.5 to 10 g/m3

 First 1952 

Uranium dioxide powder 10 g/m3 to 0.1 to 0.01 g/m3

 Schwendiman 1977 

Dust devil 5 g/m3

 Sinclair 1947 

Steam generating station 50 to 3000 mg/m3 Bond 1972, p. 62 

Mine working face (no controls) 500 mg/m3

 First 1952 

Mine air 0.05 to 0.5 g/m3

 First 1952 

Foundry workroom 2 to 30 mg/m3

 First 1952 

Los Angeles smog 0.5 to 50 mg/m3 Bond 1972, p. 62 

Nuisance dust 10 mg/m3 United Power Assoc. 1974 

Industrial atmosphere 0.1 to 50 mg/m3 Dennis 1976, p. 9 

Cigarette smoke (steady-state cocktail 

party) 5 mg/m3 Stern 1976, p. 157 

Ambient atmosphere 0.05 to 1 mg/m3 Dennis 1976, p. 9 

Air conditioned building 0.3 mg/m3

 First 1952 

Cigarette smoke (average) 40 to 400 μg/m3 Stern 1976, p. 157 

Source: Sutter 1982, Table 2.1-2 

Yu et al. (1993) also presents a review of the literature on dust loadings. Table 4.8 summarizes 

those studies. 

Table 4-8. Summary of Dust Loading Studies Cites by Yu et al. (1993) (g/m3

) 

Setting Dust Loading Author 

Urban outdoors 3.3E-5 to 2.54E-4 Gilbert et al. 1983 

Nonurban outdoors 9E-6 to 7.9E-5 Gilbert et al. 1983 

Construction activities 6E-4 Oztunali et al. 1981 

Construction traffic on unpaved roads 4E-4 Oztunali et al. 1981 

Agricultural-generated dust 3E-4 Oztunali et al. 1981 

Maximum dust loading in a cab of heavy construction 

equipment during a coal mining operation 1.8E-3 Oztunali et al. 1981 

Upper bound values report 1.3 Yu et al. 1993 

In addition, experience gained in various industries involved in the handling of bulk material, 

such as sand, coal, coke, alumina, borax, phosphate ore, and vermiculite, reported average dust 

loadings ranging from about 0.3 to 4 mg/m3

 outdoors, with peak dust loadings of up to 80 mg/m3 

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(Rando et al. 2001, Heederik et al. 1994). For workers in the concrete industry (blasting, 

drilling, grinding), the 8-hour time-weighted average dust loading of respirable particles was 

measured at between 0.26 to 14 mg/m3

 (Linch 2002). 

NIOSH should evaluate the dust-loading approach, using an average work-year dust loading on 

the order of perhaps 1 mg/m3 . Furthermore, Carlton (1999) used this dust-loading factor in an 

evaluation of dose from the Par Pond dam repair. When using this approach, the average soil 

concentration should be determined over a large work area on the order of several acres. 

4.7.2 Applicability to Revision 03 

Changes in the environmental section in Revision 03 are limited to the inclusion of dose from the 

consumption of foodstuffs. They did not resolve issues associated with dispersion models and 

resuspension factors. 

4.7.3 Summary of Overall NIOSH Response 

NIOSH believes that the dispersion model has been adequately validated by comparison to 

measured values of tritium and iodine. It should also be noted that the measured data represent 

integrated measurements that include inversion conditions when they existed. 

The comparisons made in Sections 5.1 and 5.2 of Revision 03 of the TBD also demonstrate that 

the dispersion model is claimant favorable when compared to the measured data for all areas 

modeled. In addition, because the simplified Gaussian model does not include losses due to wet 

or dry deposition, the calculated air concentrations for plutonium and uranium should also be 

claimant favorable. 

With regard to the resuspension factor, Till and Meyer 1983 chose 1×10-9 as the value most 

representative of undisturbed field conditions at SRS. NIOSH does not agree that this value is 

too low for the majority of the area over which the contamination was deposited in 1955 

(F-Area) and 1969 (H-Area), which is highly vegetated and undisturbed. 

4.7.4 Work Group Action 

The following actions were identified in Work Group deliberations: 

NIOSH Actions 

•	 Provide the Board with a written evaluation of the existing dispersion model, focusing on 

episodic releases, particularly at tank farms (e.g., contaminated solvent burns). 

4.7.5 Closure Status 

The method used to reconstruct doses to unmonitored outdoor workers due to airborne emissions 

employs an atmospheric dispersion model, assumptions, and a resuspension factor that does not 

appear to be claimant favorable and are not entirely appropriate for this class of problem. 

Revision 03 (Scalsky 2005) and proposed Revision 04-E (Scalsky 2006) provide no clarification 

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on these issues. In its response to SC&A’s finding, NIOSH maintains that the dispersion model 

has been adequately validated and is sufficiently conservative. However, the TBD has not 

specifically addressed ground-level plumes from open-pan burning of contaminated solvents or 

from environmental spills and leaks. Furthermore, NIOSH has not provided a written evaluation 

of the existing dispersion model with a focus on episodic releases, as proposed by the Work 

Group. 

The TBD does not provide the basis for the use of a resuspension factor of 1×10-9 for 

resuspension of contaminated soil. Based on an SC&A review of the literature, it also appears 

that the TBD resuspension factor of 1×10-9 per meter may not be claimant favorable. Kennedy 

and Strenge (1992) reported resuspension factors from approximately 1×10 -11 to 1×10-2 m-1 , 

which suggests that resuspension is a complex process involving several parameters and that the 

specific conditions present at the time of measurement are critical. Based on recommended 

resuspension factors presented in the literature, an average value closer to 1×10-5 to 1×10-6 per 

meter would seem more appropriate for use in worker dose reconstruction, resulting in worker 

inhalation doses from resuspension that are 3 to 4 orders of magnitude greater than those derived 

in the site profile. The dust-loading approach should also be considered, using an average workyear dust loading on the order of perhaps 1 mg/m3 . It is recommended that due to the remaining 

issues associated with the atmospheric dispersion model, assumptions, and a resuspension factor, 

this overall finding remains open. 

4.8	 COMMENT 8: INADEQUATE EVALUATION OF SPECIAL TRITIUM 

COMPOUNDS 

SRS has been one of the key facilities for the production of tritium for the U.S. defense program. 

WSRC is currently the only source of tritium in the DOE complex. The key processes at SRS 

leading to occupational tritium exposure included the following: 

• Production reactor operations 

• Product recovery in separations 

• Tritium recovery and processing 

• Laboratory research 

• Heavy water rework 

The exact forms of tritium encountered at SRS are not available; however, over the course of 

operations, SRS has handled tritium in the form of tritiated water (HTO), tritiated gas (HT), 

organically bound tritium (OBT), and metal tritides (MTs). A majority of the tritium handled 

was in the form of HT or HTO. HTO was also produced by exposure of HT and some STCs to 

air. Tritium-handling operations, research and development activities, and tritium facility 

decontamination and decommissioning may have exposed workers to other forms of tritium. 

Tritium-handling operations can form other tritium compounds, such as OBT and stable MTs. 

Organically Bound Tritium 

In a special study evaluating effluents from the tritium-handling facilities, less than 1% of the 

total tritium released was OBT in six cases out of seven. Carbon sources in the tritium 

processing area can contribute to the production of organic tritium forms by hydrogenation or 

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exchange reactions. The processes where this might have occurred include the following 

(Milham and Boni 1976): 

• Graphite crucibles used in lithium-aluminum target preparation 

• Polyethylene film from wrapping extraction crucibles 

• Ink marking for identifying targets 

• Carbon dioxide in the extraction furnace 

• Carbon dioxide in the recovery system 

• Neoprene o-ring seals throughout the process 

• Vacuum pump oil 

Milham and Boni (1976) indicated that the process associated with by-product helium 

purification produced a gaseous form of tritium that was mostly OBT (Milham and Boni 1976). 

Other tritium compounds such as rust, pump oil droplets, and tritiated solvents are known to 

form during operations with significant amounts of HTO and HT present. Observed OBT in the 

environment may be present as a result of conversion of HTO to OBT or HT to OBT (via HTO) 

in soil, plants, and animals, or from direct releases of OBT from the site. Sweet and Murphy 

(1982) documented that tritium in the soil and leaf litter near the chemical separations facility 

formed “bound” tritium as a result of the update of molecular tritium (HT) by living pine 

needles. Depending on the compound formed, concentrations of OBTs can be equivalent to or 

greater than that of HTO. 

The TBD does not adequately address potential exposures of workers to STCs including OBT 

and stable MTs during production and decommissioning activities. With documentation existing 

indicating that OBT was released to the environment, it may be reasonable to assume that it was 

present in some working conditions. ORAUT-OTIB-0011 (Siebert 2004) provides a basis for 

estimating calculated and missed dose from urine data assuming an uptake of tritiated water and 

excludes potential intakes of STCs. While the assumption of tritium as tritiated water is 

generally claimant favorable, it is not so in specialized situations. ORAUT-OTIB-0066, 

Calculation of Dose from Intakes of Special Tritium Compounds (LaBone 2007), presents 

guidance on how to calculate the best estimate dose from urine bioassay data for OBT and stable 

MTs. This procedure does not address unmonitored workers who were exposed to these 

compounds and does not specify which processes involved potential exposure to OBTs. NIOSH 

indicates that the methodology in ORAUT-OTIB-0011 (Siebert 2004) can be used without 

modification to calculate intakes to all organs or tissues, yet it does not specifically discuss the 

difference in effective half-life between HTO and OBTs (LaBone 2007). In its response to 

SC&A on June 6, 2006, NIOSH indicates that exposure to organic forms of tritium in the tritium 

facilities was possible. Furthermore, the response indicates that significant uptakes of OBTs are 

compensated for by assuming that all tritium is HTO rather than that a fraction of the total 

tritium is HT. 

The effective dose per becquerel (Bq) intake of OBT is more than twice the effective dose per 

Bq intake of HTO. The urinary excretion rate is almost the same after the second day of 

exposure. One day after exposure, the activity concentration in urinary excretion for OBT is 

57% of the HTO activity concentration in urine. As a consequence, for the same amount 

excreted in urine in the first day, the intake of OBT would be 77% higher than for HTO. Thus 

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the effective dose calculated for each Bq excreted in urine is 4 times higher since it is due to 

OBT instead of HTO. 

Stable Metal Tritides 

Metal hydrides are formed when hydrogen gas reacts with metal. SRS actively produced these 

chemical compounds during tritium processing or as a byproduct material from research 

activities (Moxley 2002). In some cases, this was done intentionally, as described by Reed et al. 

(2002). 

The production and storage of tritium, an isotope of hydrogen gas, was 

particularly tricky, and new methods were always explored to make this task 

easier and more efficient. Since most metals will react with hydrogen under 

certain conditions, the Savannah River Laboratory explored using metals to 

manipulate isotopes of hydrogen more efficiently. This led to the development of 

metal hydrides for the processing and storage of hydrogen. Metals that react with 

hydrogen to both absorb and release the gas under the right conditions, similar to 

a sponge that can absorb and release water, are called reversible metal hydrides, 

and this class of hydrides is important for hydrogen storage and processing. 

Effective reversible metal hydrides can be made from pure palladium, titanium, or 

zirconium; or from alloys of two or more metals, such as iron and titanium, or 

lanthanum and nickel. By the late 1970s, metal hydrides were used in tritium 

operations at Savannah River. This use expanded in the 1980s, and played an 

important part in the development of the Tritium Replacement Facility that began 

operations in 1994. 

The radiological concern associated with the metal hydrides occurs when tritium reacts with 

metal and produces tritium compounds, which pose challenges in personnel and radiological 

monitoring. 

Stable MTs are encountered in tritium processing facilities and in legacy materials where 

substantial amounts of tritium were handled. These compounds are solid substances containing 

tritium that does not readily react with air or aqueous solution because the tritium is tightly 

bound to the matrix. These fine particles can easily be spread by work activities and suspended 

as airborne particulates. Particles are taken into the lungs where they dissolve and release the 

tritium from the metal. SMTs may be soluble or insoluble depending on metal substrate and 

particle size and shape. The lighter MTs (e.g., lithium tritide) degrade to tritiated water, which 

can be detected through bioassay. Heavy-metal tritides behave similar to particulates. Stable 

MTs documented in unclassified sources include titanium tritide, lithium tritide, lanthanumaluminum-nickel tritide, and tritiated mercury. Zirconium tritide was under considered for use at 

SRS (Moxley 2002). 

During the February 28, 2007, visit to SRS, SC&A interviewed two individuals knowlegable of 

tritium operations at the site regarding MTs and OBT encountered on the site. The Savannah 

River Operations Office limited the interview to an unclassified discussion but indicated that 

additional questions could be submitted for consideration. The types, quantities, locations, and 

time periods where tritides were handled could not be discussed. This information is critical to 

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the evaluation of the proposed tritide dose assignment methodology. The site experts referred 

SC&A to a report discussing tritides at SRS that would inform this review. SC&A formally 

requested this document and received it on July 30, 2007. Although the document provided 

some information regarding which tritides were handled, it primarily concentrated on the 

implementation of radiological controls in areas with tritides. SC&A subsequently requested a 

second document that may provide additional detail. 

Because SMTs are relatively insoluble and the retention of this type of tritium is longer than the 

retention of HTO, the internal dose delivered to the body is higher for some of these compounds. 

For particles of these tritides, the primary organ of concern is the lung. Some of the tritium may 

leach out in the lung fluids and then be incorporated into the body water. These particles may 

also produce OBT from contact with lung tissue, which would further complicate the metabolic 

process (DOE 2004). 

McConville and Woods (1995) discuss the challenges with determining internal dose 

from MTs: 

Tritium in the form of metal tritides particles presents a peculiar problem for the 

calculation of internal dose. Standard calculations indicate that just a few 3 to 5 

micron sized particles appears to lead to a very large dose. There are very few 

data on which calculations can be based. 

Cheng et al. (1997) determined the dissolution time for fine and coarse titanium tritide particles 

to be 33 days and 361 days, respectively. McConville and Woods (1995) demonstrated, with 

individual excretion data following tritide uptakes, that tritium excretion curves for particulate 

tritides do not follow a simple exponential curve, as is the case with HTO. In the case of the 

individuals evaluated, tritides built up for a few days followed by a more traditional elimination 

curve. 

Furthermore, SMTs present unique challenges to radiological protection programs. Routine 

workplace monitoring techniques make it difficult to differentiate between STCs and more 

common forms of tritium, such as HTO. Due to the physical and chemical behavior of STCs, 

common bioassay and dose calculation models can be ineffective. For select STCs, air 

monitoring is preferable to bioassay (DOE 2004). The ICRP Database of Dose Coefficients: 

Workers and Members of the Public (ICRP 2001) provides information on tritium in particulate 

forms (Types F, M, and S). In these cases, the default parameters of lung clearance and 

absorption are applied and the biokinetic model for tritiated water is used. Thus, the dose 

coefficients from the specific MTs should be equal to the generic Types F, M, and S if the ICRP 

recommendations are followed. 

NIOSH has included in Revision 04-E of the SRS Site Profile (Scalsky 2006) a discussion on 

exposures to MTs in tritium facilities, including derivation of default intakes. In addition, 

NIOSH has included a discussion of exposure to OBT, which requires further consideration. In 

the draft version of the TBD, NIOSH has limited its application of tritide dose assessments to 

1975 and after and has exclusively applied it to workers in the tritium facilities. According to 

SRS staff interviewed, OBTs and SMTs potentially extend back to the 1950s. 

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NIOSH must be familiar with the STCs handled, the quantities of material, the locations and 

time periods of potential exposure, and the physical behaviors of tritium compounds in the 

environment (e.g., conversion to HTO, formation of rust) to correctly characterize tritium 

exposure. Bounding techniques, as proposed for SRS, cannot be effectively developed and 

applied without some basic understanding of the compounds handled and the extent to which 

individuals were exposed. This includes exposure to tritium compounds other than HTO and HT 

outside the tritium facilities. 

With these concerns in mind, SC&A does not feel that the TBD revision adequately justifies the 

upper bound approach for exposure to tritides and OBT. A bounding dose reconstruction 

methodology must consider what tritium compounds were handled or inadvertently produced, 

what percentage of tritium exists in nontraditional forms for various operations, releases of 

tritium to the work environment, and radiological controls put in place to prevent exposures. 

NIOSH should also be cognizant of the fact that STCs are not specific to SRS but may affect 

other DOE sites (e.g., Lawrence Livermore National Laboratory, Mound). 

4.8.1 Applicability to Revision 03 

Revision 03 handles the general dosimetry of exposure from STCs as a generic issue. The 

details of potential exposure to tritides and OBTs at SRS are still applicable and were not 

resolved in Revision 03. 

4.8.2 Summary of Overall NIOSH Comments 

The nature of exposure to MTs and OBT was investigated. Tritium was stored as metal 

hydrides, also called MTs, starting in the mid 1970s in the tritium facilities 232-H, 233-H, 234­

H, and 238-H. These would have been minor sources of intake relative to the total activity as 

HTO or HT. However, intake of MTs cannot be completely ruled out, and bioassay sampling for 

particulate forms of tritium was not performed. The surface contamination limit for tritium in 

accessible spaces was 106

 dpm/100 cm2 . By far the most contamination would have been HTO, 

but a conservative assumption that 50% was MTs and application of a resuspension factor of 10-4 

m

-1, which would apply to aggressive disturbance of the surfaces, would result in the following: 

air concentrationmt = (106

 dpm/100 cm2)(0.5)(104 cm

2

/m2)(10-4 m

-1) 

air conc. = 5×103

 dpm/m3 = 2.3×103 pCi/m3 . 

The daily intake is determined by the air concentration times the breathing rate per year divided 

by 365 days/year. 

intakemt = (2.3×103

 pCi/m3)(2,400 m3/yr)/365 d/yr = 1.5×104 pCi/d. 

This intake should be assigned to workers in the tritium facilities who are monitored for tritium 

from 1975 to present; it is in addition to tritium intakes determined as HTO from urinalysis. 

Either absorption Type M or S can be assigned to be claimant favorable. (In IMBA, the user 

should choose the inorganic form of tritium, toggle the inhalation mode, and assign the 

absorption type.) This intake will typically produce annual dose to most organs of a few mrem 

and about 10 mrem to the lung. 

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Exposure to organic forms of tritium (commonly called OBT) in the same tritium facilities was 

possible. Milham and Boni (1976) describe measurements of tritium species in effluents from 

the tritium processing areas, noting that, “The concentration of tritium in the organic form is 

generally 1% of the tritium released to the atmosphere.” Even if workers inside the facilities 

were exposed to several times this concentration relative to HTO, the increase to the organ doses 

calculated from urine concentrations assuming 100% HTO was insignificant. When the 

molecular sieve beds were purged, a much higher percentage of OBT versus HTO and HT was 

measured. However, this was infrequent and the effect on organ doses is compensated by 

assuming all tritium is HTO instead of accounting for the percentage of HT in the air. Unless an 

energy employee’s file provides indications of a significant intake of OBT, assuming that all the 

tritium is HTO is considered appropriate. 

4.8.3 Work Group Action 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Verify the historic timeframe, location, and quantities of tritide compounds at SRS. 

•	 Provide the Board with the planned disposition of the issue in terms of dose estimation 

methodology. 

•	 Determine the types and quantities of STCs handled at SRS to allow for a more informed 

evaluation of the tritide dose bounding methodology. 

4.8.4 Closure Status 

Revision 03 contains inadequate information regarding the assessment of dose from STCs. In 

June 2006, NIOSH proposed a methodology for the assignment of dose from STCs, which is 

provided above. Bounding techniques, as proposed for SRS, cannot be effectively developed 

and applied without some basic understanding of the STCs handled, the quantities of material, 

the locations and time periods of potential exposure, and the physical behaviors of tritium 

compounds in the environment (e.g., conversion to HTO, formation of rust) to correctly 

characterize tritium exposure. Furthermore, NIOSH limited the application of this technique to 

the period from 1975 to present. This is in conflict with site expert statements that indicate that 

potential exposures occurred back to the late 1950s. Given the large amount of tritium handled 

at SRS in various areas and the propensity of tritium to bind with organics and metals, it is 

reasonable to assume that STCs were present at SRS in some form prior to 1975. 

In April 2007, NIOSH released ORAUT-OTIB-0066 (LaBone 2007), which provided guidance 

on the assignment of dose for OBTs and SMTs using urinalysis data. Application of this 

procedure is based on process knowledge. Proposed SRS-specific guidance assigns dose from 

tritides based on surface contamination limits rather than production information and 

surveillance data, making the basis for assumptions weak, particularly in years when engineering 

controls were not as advanced as they are today. NIOSH has provided a dose estimation 

methodology for STCs; however, it has not verified the timeframe and location where STCs 

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were handled or the types and quantities of STCs handled at SRS. An evaluation of the 

adequacy of the dose estimation methodology cannot be completed without this key information; 

thus it is recommended that the issue remains open. 

4.9 COMMENT 9: HIGH-FIVE APPROACH IS BASED ON INCOMPLETE DATA 

The high-five approach is an efficiency process used to assign internal dose to employees who 

were monitored but had results less than the limit of detection, or to employees who were not 

included in the bioassay program. The technique is primarily for the assignment of dose to 

nonmetabolic organs (Brackett 2003). Intake values are derived from an average of the highest 

five intakes identified by NIOSH. Although NIOSH has indicated the contrary, it has not 

identified the highest five intakes for all radionuclides, as its search focused on the SRS Internal 

Dosimetry Registry (IDR), which is not complete. 

4.9.1 Issue Description 

The SRS TBD (Revision 03) recommends the use of a maximizing approach for likely 

noncompensable claims for nonmetabolic or digestive system cancers (Siebert 2004, p. 85). 

ORAUT-OTIB-0001 (Brackett 2003) describes an efficiency process for estimating internal 

doses for unmonitored workers for organs that do not concentrate the radionuclides in question— 

that is, for digestive tract organs and nonmetabolic organs. The approach is also applied to 

“employees who were monitored but had no detectable activity (‘positive’) in their samples and 

to employees who were not included in the bioassay program.” This is an attempt to create an 

efficiency procedure to estimate a worst-case internal dose (except for tritium) in 

noncompensable cases (Brackett 2003, p. 2): 

To facilitate timely processing of Savannah River Site claims under the Energy 

Employee Occupational Illness Compensation Program Act (EEOICPA), cases 

were reviewed to identify those with 1) little or no apparent internal dose and 2) 

cancer of an organ that does not concentrate internally deposited radionuclides 

that might be associated with work at the Savannah River Site. The cases were 

further screened to find those that met the following criteria: 

•	 No detectable activity in vitro bioassay samples, other than H-3. 

•	 No detectable activity in chest counts. 

•	 No detectable activity in whole body counts other than Cs-137, Co-60, or 

Eu-152. 

When this technique is applied to nonradiological workers and minimally exposed workers, the 

resulting internal dose is an overestimate of the actual internal dose received by these 

individuals. However, the question of whether one or more groups of unmonitored workers were 

not in either category remains to be investigated. For instance, if trade workers were 

unmonitored even when they were in hazardous job locations, the issue of onsite doses becomes 

far more complex. For those workers who were on a monitoring program and/or had the 

potential to receive internal dose, it is unclear whether the high-five approach bounds the internal 

dose. 

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The high-five approach is based on SRS intake data taken from the SRS IDR, as follows: 

These hypothetical intakes were based on recorded internal doses at SRS and 

were assumed to be composed of the radionuclides contributing the majority of 

the recorded internal dose at the Savannah River Site, except for tritium 

(assignment of tritium dose is discussed at the end of this paper). All recorded 

inhalation intakes in the history of the site were reviewed and the largest intakes 

for each radionuclide were selected. Hypothetical inhalation intakes, based on 

the recorded intakes, were used to calculate organ doses for each calendar year 

from the start of employment at the Savannah River Site through the year of 

cancer diagnosis (Brackett 2003). 

The registry includes individuals at the site who had uptakes of radioactive material and met the 

criteria applied for inclusion. The purpose of the IDR is to ensure appropriate followup bioassay 

of individuals and to ensure that workers with significant intakes are informed about the 

U.S. Transuranium and Uranium Registries (WSRC 2001). From 1951–1983, the site 

implemented the methodology from ICRP 2, Report of Committee II on Permissible Dose for 

Internal Radiation (ICRP 1960), for intake determination. An action level concentration was 

defined for each radionuclide, which was based on a fraction of a body burden. If the 

individual’s urine had a concentration in excess of the action level for that particular 

radionuclide, a follow-up bioassay sample was requested. If the second bioassay sample was 

positive, the individual was identified as having a confirmed assimilation. The SRS IDR is 

supposed to include these individuals. In about 1984, the site implemented the ICRP 30 

methodology for dose calculation for radionuclides other than tritium. At that time, the criterion 

for inclusion in the IDR was changed to include those individuals who received 100 mrem 

during the first year following intake (DPSOP 1987). With the release of the DOE Radiological 

Control Manual (DOE 1994), the criterion was changed to 100 mrem CEDE. Most recently, the 

criterion for inclusion in the SRS IDR was established at 10 mrem CEDE. The IDR includes the 

date of intake, the radionuclide, the intake quantity (nCi), annual doses for 2002 and 2003, 

CEDE, and personal information about the individual. The IDR includes intakes of Am-241, 

Cm-244, Co-60, Cs-137, Np-237, Pu-238, Pu-239, Pu-241, Sr-90, U-234, U-235, U-238, Ce-144, 

Cf-252, Cm-242, Nb-95, Ru-106, Zn-65, and Zr-95. Approximately 1,100 individuals are 

included in the registry. 

ORAUT-OTIB-0001 (Brackett 2003) used the average intake quantity for each radionuclide 

listed in the registry of the five largest documented intakes, or the average of all intakes if there 

were fewer than five intakes reported for each radionuclide (Brackett 2003). Table 4-9 

summarizes the radionuclides, intake quantities, and IREP radiation types. 

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Table 4-9. Average Hypothetical Intake Quantities for the High-Five Approach 

Radionuclide No. of 

Intakes 

Years of 

Intakes 

Average Intake (nCi) Assumed Radiation Type 

for IREP 

Am-241 5 1961–1984 14.2 Alpha 

Ce-144 5 1967–1986 623.96 

Cf-252 3 1972–1998 7.28 

Cm-242 2 1972–1979 13.35 

Cm-244 5 1964–1971 91.8 Alpha 

Co-60 1 1985 430 Electron >15 keV 

Cs-137 5 1967–1979 361.5* Electron >15 keV 

Nb-95 1 1983 650 

Np-237 5 1961–1999 1.17 Alpha 

Pu-238 5 1967–1981 250 Alpha 

Pu-239 5 1956–1979 129.1 Alpha 

Pu-241 5 1956–1979 1863.9 Alpha 

Ru-106 5 1967–1983 306.6 

Sr-90 5 1959–1986 158.18 Electron >15 keV 

U-234 5 1959–1971 105.4 Alpha 

U-235 2 1985–1990 0.14 

U-238 5 1953–1971 20.95 Alpha 

Zn-65 1 1989 700 

Zr-95 5 1967–1983 359.72 

* The intake used for this dose calculation was slightly larger than the calculated average. Because it was 

claimant favorable and had a small impact on overall dose, the calculations were not redone. 

Source: Brackett 2003 

The data used as the basis for the high-five approach are incomplete and do not capture all 

intakes that have occurred at SRS. ORAUT-OTIB-0001 (Brackett 2003) incorrectly states, “All 

recorded inhalation intakes in the history of the site were reviewed and the largest intakes 

for each radionuclide were selected” [Emphasis Added]. As mentioned above, the criterion for 

inclusion in the IDR has changed over time. While the IDR contains many of the intakes that 

occurred at SRS, neither the TBD or in ORAUT-OTIB-0001 (Brackett 2003) evaluated the 

completeness of the registry. The site profile and procedure failed to consider other sources of 

information on internal dose monitoring and incidents. Table 4-10 summarizes examples of 

intakes that potentially meet the criterion for inclusion in the high-five approach but are not 

included. Radionuclides not considered in the high-five approach identified in other source 

documents are also included. 

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Table 4-10. In-Vivo or Bioassay Information of Relevance Not Included in the High-Five 

Data 

Radionuclide Date Result 

Source: Visitor Card (i.e., 3” x 5” cards) in vivo count data 

Cm-244 8/11/72 30 nCi 

Cm-244 1/15/71 ~43 nCi 

Cm-244 2/18/72 33 nCi 

Cm-244 11/6/72 44 [31] nCi 

I-131 6/30/72 2.6 nCi 

Pu-239 2/29/72 64 nCi 

Pu-239 8/24/72 63 nCi 

Pu-239 8/23/72 117 [103] nCi 

Pu-239 6/21/72 57 [49] nCi 

Cs-134 11/9/70 85 nCi 

Cs-134 2/22/71 1.34 nCi 

Cs-134 11/9/70 1.9 nCi 

Sb-122 11/9/70 206 nCi 

Sb-124 11/9/70 78 nCi 

Sb-124 1/15/71 283 nCi 

Source: Bioassay Results in DPSP Monthly Reports* 

Pu-239 May-66 5.0 dpm/1.5 L 

Pu-239 Jul-57 8 dpm/1.5L 

Pu-239 Feb-62 9.38 dpm/1.5L 

Pu-239 Mar-61 9.6 dpm/1.5L 

Pu-239 Jun-60 9.8 dpm/1.5 L 

FP Aug-56 1220 dpm/750mL 

FP Oct-56 3530 dpm/750mL 

FP Jan-57 1628 dpm/750m 

FP Oct-59 567 dpm/750mL 

FP Jun-59 413 dpm/750mL 

U-235 Sep-57 320 dpm/1.5L 

U-235 Oct-57 146 dpm/1.5L 

U-235 Dec-57 232 dpm/1.5L 

U-235 May-57 15 dpm/100mL (225 dpm/1.5 L) 

U-235 Feb-61 85 dpm/1.5L 

U-235 Nov-60 24 dpm/1.5L 

Np-237 Feb-61 0.24 dpm/1.5L 

Np-237 Oct-61 4.13 dpm/1.5L 

Np-237 Feb-63 0.5 dpm/1.5L 

Np-237 May-63 4.0 dpm/1.5L 

Np-237 Aug-65 0.3 dpm/1.5L 

Total U Nov-59 33 ug/L 

Total U Apr-61 3 ug/L 

Total U May-66 41 ug/L 

Total U Apr-64 48 ug/L 

Total U Dec-56 52 ug/l 

*Note: The dates listed represent the date of the report, not the bioassay sample. 

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The above examples are only a fraction of the data identified from the DPSP monthly progress 

reports. Figure 4-2 depicts a synopsis of the bioassay data in the monthly reports versus those 

from individuals included in the high-five analysis for Pu-239 and Np-237. The monthly reports 

generally provide bioassay information up through 1965 that is higher than that observed for 

individuals represented in the high-five analysis. The monthly progress reports accessible to 

SC&A were available only through 1965 for this analysis. 

Figure 4-2. Pu-239 and Np-237 Bioassay Comparison between Data in the Monthly 

Reports and Data Used for the High-Five Approach 

Progress reports also included generic statements of relevance that warrant further investigation. 

The progress report from the Works Technical Department for December 1960 (DPSP 1960k) 

indicates that other intakes of Nb-95 occurred prior to 1983: 

Routine examination of manufacturing area personnel began early in December. 

The first employees scheduled were from the Health Physics Section in the 

Separations Areas. Body burdens of 58 men were measured. Ce-Pr144

 was 

detected in 66% of the personnel, Ru-Rh106 in 29% and Zr-Nb95 in 22%. The 

maximum body burdens found in these individuals were 49 nanocuries of 

Ce-Pr144, 36 nanocuries of Ru-Rh106, 7.2 nanocuries of Zr-Nb95 and 24 nanocuries 

of Cs-Ba137 . 

To further bolster the assertion that uranium posed unimportant risks, McCarty (2000) states the 

following: 

These protection measures, not withstanding, records indicate that 99 workers 

received internal doses of uranium over the history of the plant, which are well 

documented in site incident reports. 

There is concern that this number of uranium uptakes is based on data currently being used by 

ORAU for dose reconstruction purposes. However, a preliminary review of an incomplete set of 

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Works Technical Department reports indicates that there were 155 positive bioassays for 

uranium between 1953 and 1959 alone.3 

There are also other indications that high intakes, possibly higher than those listed, may have 

been missed. Only one of the five entries for Pu-239 in Table 1 of ORAUT-OTIB-0001 

(Brackett 2003) regarding the largest intakes assigned at SRS (based on radionuclides available 

in IMBA) is from the 1950s; one is from 1962 and the rest are from the 1970s (Brackett 2003, p. 

4). This is surprising since the highest exposures in the nuclear weapons complex typically 

occurred in the 1960s and earlier. None of the high Cs-137 intakes in Table 1 are from the 

1950s. Three of the high Sr-90 intakes are reported on the same day in the 1980s (November 5, 

1986) and are very close in value. As another example, the production of Pu-238 from Np-237 

targets started in the late 1950s and ended in 1986 (Reed et al. 2002, p. 429), but four of the five 

Np-237 entries in Table 1 are from the 1990s. At the same time, all of the Pu-238 entries are 

from the period of production. These are among the indicators that the records used to compile 

the high-five analysis may be inadequate to determine the highest five intakes for the listed 

radionuclides. Table 2 of ORAUT-OTIB-0001 (Brackett 2003) does not contain any entries for 

the 1950s for fission products. 

SC&A’s review also revealed that incidents during the early years may have been underreported. For example, at the time of a significant incident, one would expect a follow-up that 

included a detailed review of the circumstances and special bioassay monitoring of the exposed 

individual. These types of follow-up activities likely occurred, because SC&A noted more than 

400 cases of post incident chelation therapy. Based on these findings, SC&A suggests that the 

site profile provide direction to dose reconstructors on how to identify the occurrence of 

incidents, investigate incidents, and obtain records and data sources that can be useful in 

reconstructing doses from incidents when bioassay data or personnel dosimetry are lacking or 

suspect. 

Several radionuclides had no bioassay technique available in the early years during peak 

production. For example, the TBD indicates that analytical methods for Np-237 were not 

available until about 1959, and that analytical methods for americium, curium, and californium 

were not available until the mid-1960s. One of the first incidents investigated at the site in 

September 1954 involved the spread of contamination from an americium source (Nichols et al. 

1954), demonstrating that americium was present on the site prior to the development of a 

bioassay technique. The lack of monitoring data from this era of operations also brings into 

question whether the highest intakes for each radionuclide have actually been captured. 

The TBD does not mention Am-243, Eu-152, Eu-154, M-54, Sb-125, Th-228, Th-232, and U236, although the SRS Internal Dosimetry Technical Basis Manual (WSRC 2001) lists them as 

radionuclides of concern. ORAUT-OTIB-0001 (Brackett 2003) outlines the default radiation 

type for radionuclides in the high-five analysis. It is unclear why NIOSH restricts radionuclides 

3

 DPSP 1956, DPSP 1956a, DPSP 1956b, DPSP 1956c,, DPSP 1956e, DPSP 1956f, DPSP 1957a, DPSP 

1957b, DPSP 1957c, DPSP 1957d, DPSP 1957e, DPSP 1957f, DPSP 1957g, DPSP 1957h, DPSP 1957i, DPSP 

1957j, DPSP 1957k, DPSP 1958a,, DPSP 1958b, DPSP 1958c, DPSP 1958d, DPSP 1958e, DPSP 1958f, DPSP 

1959a, DPSP 1959b, DPSP1959c, DPSP 1959d, DPSP 1959e, DPSP 1959f, DPSP 1959g, DPSP 1959h, DPSP 

1959i, DPSP 1959j, DPSP 1959k, DPSP1959l. 

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to only one radiation type when multiple high-yield radiation types are associated with a 

radionuclide. For example, NIOSH does not consider photon exposures from Cs-137 or Co-60. 

Radiation types should represent all high-yield forms emitted by the particular radionuclide. 

In summary, the completeness of the IDR should be evaluated against incident and technical 

reports, as well as against operations history. NIOSH should also evaluate the adequacy of the 

bioassay program during the period of operation and the radionuclides in the source terms with 

regard to the complex history of radionuclide production. Furthermore, NIOSH should consider 

the impact of the bioassay program’s adequacy on internal dose reconstructions. The high-five 

approach cannot be used to cover all omissions of data for internal dose since its use is limited to 

organs that do not concentrate internally deposited radionuclides and to workers with little or no 

apparent internal dose. 

4.9.1.1 3"x 5" Visitor Cards 

In the past, the 3"x5" cards were used to record internal (i.e., in vivo) and external monitoring 

data for visitors, subcontractors, construction workers, and employees who were assigned a 

temporary badge. The 3"x5" cards are sometimes referred to as visitor cards. Historically, these 

cards were not typically included in the Health Physics files, but were stored separately. SRS 

scanned the data from all the 3"x5" cards into pdf format (Morgan 2006). 

The NIOSH Health Related Energy Research Branch (HERB) obtained imaged copies of at least 

some of the visitor cards for the leukemia case-control study. OCAS was referred to NIOSHHERB for a copy of these 3"x5" cards, but was unable to locate them or obtain any copies. 

SC&A provided NIOSH-OCAS with a few examples of these cards on June 29, 2007, to assist 

with their search. In lieu of these card images, NIOSH reviewed 20 SRS claims with early work 

histories for these cards and located scanned data cards within claimant files. NIOSH further 

noted that the information on the cards only related to external dose. However, in a review of a 

fraction of 3” x 5” cards, SC&A noted that a few cards also contained internal monitoring data. 

NIOSH should verify that all 3”x5” card data is included in the claimant files submitted by SRS, 

and that both internal and external monitoring information from these cards are considered 

during dose reconstruction. 

SRS entered data after 1970 into the Health Visitor Information System (HVIS) database. It is 

uncertain how many of these data were entered into HPAREH or HVIS in the version used by 

NIOSH. SC&A could provide NIOSH with a limited subset of this data; however, it is more 

beneficial for NIOSH to request the complete set of data from SRS. 

4.9.2 Applicability to Revision 03 

This issue is applicable to Revision 03 and ORAUT-OTIB-0001 (Brackett 2003). Revision 03 of 

the site profile provides the same direction to dose reconstructors as the previously reviewed 

version. ORAUT-OTIB-0001 (Brackett 2003) describes the high-five methodology in detail. 

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4.9.3 Summary of Overall NIOSH Response 

It is extremely unlikely that data for larger intakes exist yet were not included in the database 

used. The specific purpose of the database was to capture all significant intakes at SRS. The 

high-five approach includes only the largest five intakes for any given nuclide. 

SC&A provided an email regarding the 3"x5" cards. NIOSH asked ORAUT on October 10, 

2006, to determine whether these 3"x5" cards already existed in the HERB data folder on the odrive. No 3"x5" cards exist in the HERB data folder on the o-drive, nor does NIOSH have 3"x5" 

cards in any of its records collections. In July 2007, NIOSH made a special data capture trip to 

SRS to attempt to discover more information about intakes. The trip yielded standard bioassay 

cards, not 3"x5" cards but instead 6"x9" cards. NIOSH therefore asked SC&A to clarify the 

issue. 

NIOSH indicated that the original text file of bioassay data for those individuals included in the 

high-five approach analysis was located on the O-drive at O\AB document review\SRS TBD 

review\srs.text. 

4.9.4 Work Group Action 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Identify the location of confirmed intake registry and bioassay data on the o-drive. 

(completed) 

•	 Complete the update of high-five intakes using new models and provide this to the Board. 

•	 Review the incident database (i.e., Fault Tree Data Bank) for applicability. 

•	 Make the copy of the incident databases available for Work Group review when the copy 

is received. 

SC&A Actions: 

•	 Assist in providing additional information on the SRS 3"x5" visitor cards and identify the 

intake information discussed during Work Group meetings. (Completed) 

4.9.5 Closure Status 

Revision 03 includes the same information on the high-five approach as the previous revision 

reviewed by SC&A. Proposed Revision 04-E (Scalsky 2006) clarifies the purpose and 

application of the high-five approach but does not resolve the issues related to the basis for the 

high-five approach. The completeness of the data used as the basis for the high-five approach is 

questionable. NIOSH has not reviewed all recorded inhalation intakes in the history of the site 

despite its indication to the contrary. The criteria for inclusion of individuals in the SRS intake 

file have changed over time such that some intakes are excluded. Other sources of information 

such as the DPSP monthly reports, data banks, incident files, and visitor cards were not 

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considered when identifying intakes. SC&A has presented numerous examples from these 

sources that likely meet the criteria for inclusion in the high-five approach and deserve further 

investigation. The general bioassay trend for Np-237 and Pu-239 from the individual high-five 

data and the DPSP monthly report data indicates that monthly report bioassay results were 

higher. In addition, some radionuclides were present at the site prior to the availability of 

bioassay techniques. It is recommended that several Work Group actions remain open, including 

clarification of the location of the SRS high-five bioassay data, review of the tank farms Fault 

Tree Data Bank, and completion of the revised high-five approach. 

4.10	 COMMENT 10: TECHNICAL ISSUES ASSOCIATED WITH THE HIGH-FIVE 

APPROACH 

Technical issues are associated with the high-five approach. The use of surrogate data for 

internal dose for unmonitored workers and for target organs that do not concentrate the 

radionuclides in question is not necessarily a maximizing approach for making dose estimates, 

contrary to the claim in the TBD. 

4.10.1 Issue Description 

The hypothetical intake outlined in ORAUT-OTIB-0001 (Brackett 2003) uses the SRS IDR to 

identify the highest five intake quantities (nCi) for each radionuclide in the IDR, or all available 

intakes if the reported intakes for a given radionuclide are less than five. The intake quantities 

calculated using ICRP 30 methodology is then averaged. The average activity (nCi) is entered 

into IBMA and a dose is calculated based on the ICRP 66 and ICRP 68 models. For each, the 

dose reconstructor is instructed to assume an acute inhalation occurred on January 1 in the first 

year of employment (Brackett 2003). 

NIOSH justified the use of intakes calculated with the ICRP 30 methodology rather than the 

newer ICRP methodology (e.g., ICRP 66, ICRP 68) prescribed for the dose reconstruction effort 

by comparing intake retention fractions (IRFs) from ICRP 30 and ICRP 68. This justification is 

not necessarily claimant favorable. The use of ICRP 30 models does not produce intake values 

that are higher than those derived by the new ICRP models for a majority of the relevant 

radionuclides included in the hypothetical intake. 

Plutonium and americium are not significantly overestimated as stated in ORAUT-OTIB-0001 

(Brackett 2003, pg. 8). In fact, intakes from Zr-95 (Types M and F), Zn-65 (Type S), Ru-106 

(Type S), Nb-95 (Type M), Cf-252 (Type M), Ce-144 (Type M), Cs-137 (Type M), Co-60 (Type 

M), Sr-90 (Type S), uranium (Types F, M, and S), plutonium (Types M and S), and Am-241 

(Type S) for all reasonable times of collecting samples, after an intake occurred, are 

underestimated using ICRP 30 methodology instead of the ICRP 68 biokinetic model. Types M 

and F of Ru-106 are underestimated most of the time using ICRP 30 methodology. For Type M 

of Am-241, ICRP 30 methodology may or may not underestimate the intakes, depending on the 

time samples are taken after the intake. Attachment 6 of SC&A’s previous review (SC&A 

2005a) presents calculations demonstrating that the approach used in ORAUT-OTIB-0001 

(Brackett 2003) is not claimant favorable for many radionuclides and that ICRP 68 models 

would have been more claimant favorable. 

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4.10.2 Summary of Overall NIOSH Response 

This approach is used as an overestimate for people who were not monitored or had no positive 

bioassay results. It assigns all possible nuclides for which there are recorded intakes on the site 

and treats isotopes of the same nuclide independently. For example, an individual is assigned the 

largest U-238 intake, the largest U-234 intake, and the largest U-235 intake, even though these 

would have been from different source terms (depleted versus enriched uranium). It is not 

plausible that an unmonitored individual would have a larger intake of every radionuclide than 

the largest recorded in the history of the site without some kind of indicator. 

The data for the individual workers upon which the high-five doses are based have been 

retrieved from SRS. These data are currently being entered into an electronic database. The 

reevaluation of the cases using the newer ICRP (60, 66–69, 71–72) models is expected to be 

completed by the end of August 2006. 

4.10.3 Work Group Action 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

• Complete the update of high-five intakes using new models and provide this to the Board. 

4.10.4 Closure Status 

Revision 03 (Scalsky 2005) and Revision 04-E (Scalsky 2006) reference ORAUT-OTIB-0001 

(Brackett 2003), which continues to use both ICRP 30 and subsequently ICRP 66 as the method 

for assignment of internal dose. ORAUT-OTIB-0001 (Brackett 2003) justifies the use of intakes 

calculated with the ICRP 30 methodology rather than the most current ICRP methodology by 

comparing IRFs from ICRP 30 and ICRP 68. The ICRP 30 model does not produce intake 

values that are higher than those derived by the new ICRP models for a majority of the relevant 

radionuclides included in the hypothetical intake. The use of the ICRP 30 methodology to 

calculate the intake, with a subsequent use of ICRP 68 models to calculate the dose, did not 

always result in the intended highest dose to an organ. Similarly, the appropriate solubility types 

between the two methodologies were not always paired consistently, resulting in discrepancies 

that were not claimant favorable. It is recommended that this issue remain open pending the 

update of the high-five method using bioassay data and appropriate models. 

4.11 COMMENT 11: ICRP 30 WITH SUBSEQUENT USE OF ICRP 68 

For internal dose calculations, the use of ICRP 30 methodology to calculate the intake, with a 

subsequent use of ICRP 68 models to calculate the dose, did not always result in the intended 

highest dose to an organ. Similarly, the appropriate solubility types between the two 

methodologies were not always paired consistently, resulting in discrepancies that were not 

claimant favorable. 

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4.11.1 Issue Description 

The hypothetical intake outlined in ORAUT-OTIB-0001 (Brackett 2003) uses an average of the 

highest five intakes in the SRS IDR quantities calculated using the ICRP 30 methodology. The 

average activity (nCi) is entered into IBMA and a dose is calculated based on the ICRP 66 and 

ICRP 68 models. For each dose reconstruction where bioassay data are lacking (and which meet 

certain other criteria that are described below), the dose reconstructor is instructed to assume that 

an acute inhalation occurred on January 1 in the first year of employment (Brackett 2003). The 

use of two different ICRP methodologies requires assumptions of two solubility types. The first 

solubility assumption was made by SRS when they calculated the intake using ICRP 30. The 

second solubility assumption was made by NIOSH when the intake was entered into IMBA. 

The appropriate solubility types applied in the comparison were not always paired between the 

two methodologies (i.e., Type F corresponding to Class D, Type M corresponding to Class W, 

and Type S corresponding to Class Y). Instead, the ICRP 68 solubility types are chosen “as the 

most soluble form of the material because it would clear from the lung more rapidly than 

insoluble material, thus depositing in the organ or tissue sooner” (Brackett 2003). The ICRP 30 

classes, on the other hand, are chosen using “the material class(es) applied for the SRS calculated 

intakes in Tables 1 and 2” (Brackett 2003, pp. 5–8). There is a fundamental problem with the 

comparisons of these IRFs from ICRP 30 and ICRP 68. In general, when intakes are used to 

calculate organ doses, the choice of the most soluble type is claimant favorable for doses 

calculated to systemic organs. When bioassay results are used to calculate organ doses, many 

times the assignment of the most insoluble material type results in a higher dose for systemic 

organs, as illustrated by the following example (SC&A 2005a, p. 152): 

•	 A 24-hour urine sample is collected five days after a single inhalation 

intake of Pu-238. The bioassay result is 1 Bq of Pu-238. Using the model 

in Age-dependent Doses to Members of the Public from Intake of 

Radionculides: Part 2, Ingestion Dose Coefficients (ICRP 67) for 

plutonium, the calculated intakes are as follows: 

- For Pu-238, Type S: intake of 2.2×106 Bq [50-year committed bone 

surface dose is 75 sieverts (Sv), 50-year committed dose to colon is 

0.053 Sv, 1-year committed dose to the colon is 0.006 Sv] 

- For Pu-238, Type M: intake of 2.6×104 Bq (50-year committed bone 

surface dose is 24 Sv, 50-year committed dose to colon is 0.042 Sv, 

1-year committed dose to the colon is 0.002 Sv) 

Thus, the use of Pu-238, Type S, results in a higher intake than the use of Type M (and in 

higher doses to systemic organs). 

•	 Using the ICRP 30 IRF from Table 3 of ORAUT-OTIB-0001 (Brackett 

2003) , the same bioassay result of 1 Bq of Pu-238 in a 24-hour urine 

sample taken 5 days after a single intake corresponds to the following 

intakes: 

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- For Class Y: intake of 3.5×105 Bq (ICRP 30) (50-year committed 

bone surface dose is 12 Sv, 50-year committed dose to colon is 

0.33 Sv, 1-year committed dose to the colon is 0.037 Sv) 

- For Class W: intake of 1.9E4 Bq (ICRP 30) (50-year committed bone 

surface dose is 17.5 Sv, 50-year committed dose to colon is 0.03 Sv, 

1-year committed dose to the colon is 0.0015 Sv) 

Thus, the more claimant-favorable approach to choosing solubility type should be initiated with 

the intake calculation and not limited to the internal dose calculation. 

ORAUT-OTIB-0001 (Brackett 2003) directs the dose reconstructor to use surrogate 

radionuclides for radionuclides included in the high-five approach, which are not available in 

IMBA. Since the initial review of the SRS site profile, the IMBA code was updated to include 

radionuclides previously not available, eliminating the need to use surrogate organs. ORAUTOTIB-0001 (Brackett 2003) should be updated to reflect this change. 

4.11.2 Applicability to Revision 03 

This comment is applicable to Revision 03 and ORAUT-OTIB-0001 (Brackett 2003). The 

proposed use of bioassay data will resolve this issue. 

4.11.3 Summary of Overall NIOSH Response 

ORAUT-OTIB-0001 (Brackett 2003) contains information demonstrating that intakes calculated 

using ICRP 30 IRFS are in most cases larger than those obtained from ICRP 68 IRFs. The 

subsequent dose would have been smaller had ICRP 30 DCFs been used, but ICRP 68 DCFs 

were used. As noted in the response to Comment 10, NIOSH did not have access to the original 

bioassay results at the time ORAUT-OTIB-0001 (Brackett 2003) was written; SC&A had the 

intakes that had already been calculated by SRS. Solubility types were not chosen to match the 

SRS-calculated intakes; they were selected to maximize dose to the systemic organs once an 

intake had already been determined. 

It should also be noted that the latest version of IMBA contains radionuclides previously not 

available, eliminating the need for surrogate radionuclides in many cases. 

4.11.4 Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

• Complete the update of high-five intakes using new models and provide this to the Board. 

4.11.5 Closure Status 

Revisions 03 and 04-E of the TBD reference ORAUT-OTIB-0001 (Brackett 2003), which 

continues to use two internal dose assessment methodologies with at inappropriate pairing of 

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solubilities. Furthermore, the procedure includes recommendations for the use of surrogate 

organs. For internal dose calculations, the use of ICRP 30 methodology to calculate the intake, 

with a subsequent use of ICRP 68 models to calculate the dose, did not always result in the 

intended highest dose to an organ. Similarly, the appropriate solubility types between the two 

methodologies were not always paired consistently, resulting in discrepancies that were not 

claimant favorable. The dose reconstructor is directed to use surrogate radionuclides for 

radionuclides absent from the IMBA code. The optimum solution is to add these radionuclides 

to the code. In lieu of this, a more prudent approach to address the absence of radionuclides is to 

use the dose coefficients provided in ICRP (2001) and employ a linear interpolation for the 

radionuclides that are not explicitly given. This issue has been partially resolved since the initial 

SC&A review with the inclusion of additional radionuclides in the latest version of the IMBA 

code, eliminating the need to use surrogate organs in many cases. ORAUT-OTIB-0001 

(Brackett 2003) should be updated to reflect these and other changes to the high-five approach. 

The closure of this issue is pending the update of the high-five method using bioassay data and 

appropriate models. 

4.12	 COMMENT 12: SOLUBILITY, ORO-NASAL BREATHING, AND INGESTION 

ASSUMPTIONS 

Internal dosimetry needs to be improved with regard to radionuclide solubility, oro-nasal 

breathing, and the ingestion pathway. These factors should be carefully considered with regard 

to internal dose reconstruction. SC&A originally developed the points described below for the 

review of the Bethlehem Steel and Mallinckrodt Chemical Works site profiles, and they are 

applicable for all bioassay interpretations for EEOICPA. 

4.12.1 Solubility Assumptions 

The solubility assumptions that are used to estimate organ dose from urine need to be discussed 

in the TBD. For example, an assumption of Type S or Type M (and Type F in the case of UNH) 

must be considered more carefully when deriving doses to organs based on urinalysis data, since 

a Type S assumption in this case may yield a higher dose for nonrespiratory tract organs than a 

Type M assumption. The analysis of organ dose from urine data can be complex, and more 

specific analysis is needed in any future revisions of the TBD. 

Oro-nasal Breathing 

SC&A has addressed oro-nasal breathing in detail in Attachment 5 of SCA-TR-TASK1-0002, 

Review of NIOSH Site Profile for Mallinckrodt Chemical Company, St. Louis Downtown Site St. 

Louis, Missouri (SC&A 2005c). That finding is also applicable to SRS workers. Oro-nasal 

breathing affects intakes for light as well as heavy work. The assumption of oro-nasal breathing 

should be used in a manner similar to solubility assumptions—that is, uncertainty as to whether a 

worker breathed through the mouth should be addressed and a determination made whether 

NIOSH should continue to strictly follow ICRP models, which do not address oro-nasal 

breathing, or whether oro-nasal breathing should be included in dose reconstructions as a more 

claimant-favorable assumption. Oro-nasal breathing needs to be taken into account when air 

concentration data are used in estimating intakes and doses, as for instance in estimating 

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environmental occupational dose (Section 3 of the TBD) or missed dose due to fission products 

(p. 80 of the TBD). 

Ingestion 

NIOSH/ORAUT has assumed that inhalation is the only pathway for internal exposure at SRS. 

During discussions between SC&A and NIOSH/ORAUT, the following issue arose with regard 

to ingestion doses: 

Question: 

For purposes of internal dose calculations; are airborne release levels well 

documented; are potentials for ingestion and inhalation sufficiently documented; 

are bioassay techniques well documented and is each bioassay technique’s 

uncertainty and accuracy well understood? 

Answer: 

“Potentials for inhalation” are accounted for by estimating missed dose or 

accounting for unmonitored periods. Ingestion is not usually considered at major 

DOE sites [however, it is important at Atomic Weapons Employers], but uptake 

from the GI tract is accounted for in the bioassay, although the default intake 

mode is inhalation unless a worker’s records have information indicating 

otherwise. 

SC&A believes that ingestion cannot be ignored a priori by assuming a default value. 

Furthermore, in order to take ingestion into account using bioassay data, the inhalation 

component must be known. In other words, a single bioassay result gives one data point, but 

there are two unknowns: how much was inhaled and how much was ingested. One cannot solve 

this problem accurately without one more data point. 

There are several ways to approach this problem. The first, of course, is to look for an additional 

data point. This could be provided by an in vivo count, for instance. The problem would also be 

solvable if fecal analysis and urinalysis data were both available for estimating the intake in 

question. Although it currently does not, the TBD would have to then specify a procedure for 

solving for the inhalation and ingestion intakes. Revision 03 of the TBD adds the annual dose 

from ingestion of foodstuffs to the environmental discussion in Section 3.0. Doses are calculated 

for the ingestion of 60 kilograms (wet mass) of food per year for Sr-90, Cs-137, Pu-238, Pu-239, 

and tritium. 

This gap creates a potential issue at all DOE sites. It should also be noted that this issue does not 

affect only the early period; it will not appear on any worker records. The potential maximum 

effect on organ doses for each important radionuclide can be assessed by a simple screening 

technique (not to be misunderstood as a dose estimate). If one assumes inhalation is equal to 

zero in interpreting the bioassay data, then one can obtain a theoretical maximum ingestion value 

and corresponding organ doses. This puts an upper limit on potential errors. NIOSH may wish 

to perform these types of screening analyses in order to close out this issue. 

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4.12.2 Applicability to Revision 03 

Revision 03 includes a discussion of dose from the ingestion of foodstuffs that was not included 

in previous revisions. 

4.12.3 Summary of Overall NIOSH Response 

Although ORAUT believes that the assumptions for internal dosimetry are reasonable, it 

understands this to be a generic issue outside the scope of the SRS site profile that is being 

addressed by NIOSH and SC&A. 

4.12.4 Work Group Actions 

There are no Work Group Actions at this time. 

4.12.5 Closure Status 

Section 3.0, Revision 03 of the TBD, now contains an annual dose from the ingestion of 

foodstuffs. Doses are calculated for the ingestion of 60 kilograms (wet mass) of food per year 

for Sr-90, Cs-137, Pu-238, Pu-239, and tritium. Proposed changes in Revision 04-E (Scalsky 

2006) do not include any additional information on ingestion dose. While inhalation is likely to 

be the predominant intake mode the vast majority of the time, the TBD needs to provide an 

overall analysis to sustain the default assumption that inhalation is always the predominant 

intake mode. The importance of such an evaluation is increased because doses are being 

estimated for individuals rather than populations. 

Onsite atmospheric dispersion and resuspension. The method used to reconstruct doses to 

unmonitored outdoor workers due to airborne emissions employs an atmospheric dispersion 

model, assumptions, and a resuspension factor that do not appear to be claimant favorable and 

are not entirely appropriate for this class of problem. Revision 03 and proposed Revision 04-E 

(Scalsky 2006) do not provide any additional clarification of these issues. SC&A recommends 

that this issue remain OPEN pending further Work Group review. 

4.13	 COMMENT 13: INCIDENTS AND HIGH-RISK JOBS NOT ADEQUATELY 

ADDRESSED 

The TBD does not list or reference incidents and high-risk jobs to alert dose reconstructors to 

unique exposure conditions. 

4.13.1 Issue Description 

The TBD does not address exposure conditions that may present themselves during an incident 

or occurrence. SC&A maintains that some workers have sustained significant exposures that are 

relevant to their dose reconstruction but may not be captured by reviewing the DOE individual 

dosimetry file and the CATI. In site expert interviews conducted during the initial review, SRS 

staff indicated that the individual dosimetry file contains many, but not all, incident 

investigations. Furthermore, SRS implemented several methods for documenting incidents and 

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occurrences. Other sources of incident information include SHI reports, the WSMS incident 

database, the tank farms data bank, Works Technical Department progress reports, and multiple 

other incident files corresponding to the main areas of the site (e.g., tritium facility incidents, raw 

materials incidents, and reactor incidents). These reports provide brief descriptions of the 

incidents and were maintained separately from the individual radiation exposure records. 

Although recommended in the original SC&A review, NIOSH only recently (following the 

February 2007 site visit) obtained a copy of the SHIs. Prior to this, these reports were not 

considered in the dose reconstructions or development of efficiency methods. In addition, the 

tank farm data bank entries have not been evaluated to identify significant exposure situations 

and environmental releases, which may be significant in dose calculations. SC&A was unable to 

obtain a copy of the tank farm data bank for this review. 

Per the telephone conversation held between SC&A and NIOSH during the previous review, the 

DOE exposure file and the CATI provide the mechanisms for identifying incidents, and the SRS 

personnel radiation exposure file includes many incident reports. The CATI interview is used as 

a secondary source of information on incidents or occurrences. While the CATI provides some 

potential to identify incidents that may be missing from the personnel radiation exposure file, 

reliance on the CATI for this purpose places an inappropriate burden upon the worker or 

survivor to recall events and to recognize potential implications for the dose reconstruction. This 

concern particularly impacts family member claimants, as they are far less likely to have 

knowledge of incidents and their relevance to the claim. As a result, there is a disadvantage for 

dose reconstructions that must rely on the CATI to determine the possibility that the worker was 

involved in an incident. The CATI should be used as a positive indicator of an incident; 

however, it should not be used to rule out the existence of incidents. 

Although individuals involved in incidents are usually monitored, the incident itself may pose 

special exposure conditions that need to be considered in the dose reconstruction (e.g., injection 

versus inhalation, partial body exposure to an external beam, cleanup of a spill involving 

nontraditional radionuclides). Incident databases maintained by SRS provide valuable 

information on the nature of incidents and source documents and reduce the disadvantage to 

family member as opposed to energy employee claimants. A redundant system for incident 

identification is necessary for an effective evaluation of incidents and accidents. The high-five 

approach cannot be used to cover all omissions of data for internal dose since its use is limited to 

organs that do not concentrate internally deposited radionuclides and to workers with little or no 

apparent internal dose. Without a thorough reconciliation of the DOE individual dosimetry files 

against these separate incident data banks, NIOSH cannot be assured that all significant 

exposures from incidents are submitted by SRS and considered in relation to individual worker 

claims or to high-five estimates of maximal doses. 

One source of incident information mentioned in the original SC&A review is the SHIs. 

DPSOP-40, Special Hazard Bulletin 2, Investigating Radiation and Contamination Incidents 

(DPSOP 1981), contains the standard operating procedure for SHIs. The following are among 

the incidents required to be investigated under this procedure (DPSOP 1981): 

- Acts or situations which caused or could have caused hazardous radiation or 

contamination conditions. 

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- Contamination incidents which can lead to significant loss of containment of 

radioactivity, require costly clean up, or concern to Health Protection. 

- Incidents that result in body contamination or radiation exposure of concern 

to Health Protection or Medical. 

SC&A’s previous review (SC&A 2005a) included several examples of incidents from the tank 

farm Fault Tree Data Bank that meet the criteria for SHIs (DPSOP 1981). Only three of the 12 

incidents from the databank were not found in DuPont’s log of SHIs (DuPont 1990). Incidents 

selected from the tank farm data bank contained three estimated internal exposure entries (one in 

the F-Area in 1972 and two during the same incident in the H-Area in 1974) that were larger than 

the lowest two values listed in ORAUT-OTIB-0001 (Brackett 2003), Table 1, for the high-five 

Cs-137 intakes. The average for the NIOSH high-five Cs-137 intake is about 361 nanocuries, 

while the SC&A average, using multiple sources of data, is about 475 nanocuries, or about 31% 

higher. Furthermore, it is unclear how the standard operating procedure for SHIs has changed 

over time. 

SRS staff obtained a copy of a database (containing major and minor incidents through 1999) 

from WSMS. Because this database was lengthy and contained classified information, a review 

team assembled on site to review its content. Sam Glover (NIOSH), Elyse Thomas (ORAUT), 

Brad Clawson (Advisory Board), Mark Griffin (Advisory Board), and Kathryn RobertsonDeMers (SC&A) visited SRS from February 28 through March 1, 2007, to review an incident 

database thought to be the tank farm Fault Tree Data Bank, and talk to individuals 

knowledgeable about this particular database. The three objectives of the visit were to determine 

the contents of the database, compare entries in this database to those from the tank farm Fault 

Tree Data Bank, and determine its usefulness in dose reconstruction. As a secondary task, the 

group conducted an interview specifically dealing with STCs at the site. 

A former SRS employee developed the WSMS database to provide data for safety analysis. It 

incorporates information from progress reports, incident investigation reports, field logbooks, 

and other sources containing safety incident information. The main table of the database, 

referred to as “incidents,” includes as core information an incident number, incident date, 

database entry date, and a text description of the event. Five definition tables identify the source 

of information, area, facility, operation, and equipment. Five intermediate tables contain 

information necessary to link the definition tables to particular incidents. The incidents table 

does not identify individuals involved in the incident, but this information may be contained in 

the referenced source documents. The review team was not able to interpret the codes used to 

identify source documents in the database and was therefore unable to pull source documents to 

examine their content. SRS was asked to provide a database manual to define the database codes 

more clearly. SRS has determined that the originator of the database is alive and has agreed to 

attempt to contact him. It would be beneficial for the reviewers to interview this individual to 

better understand the content and objectives of the database. 

The team reviewed a representative collection of incidents from the WSMS database to evaluate 

the scope and potential utility of the information. The data appear to cover a limited scope of 

site operations. Most incidents relate to the 200-F and 200-H Areas, with minimal data from the 

reactor areas, heavy water area, the Savannah River Technical Center, the raw materials area, 

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and other site operations. The vast majority of entries reviewed (more than 90%) relate to the 

separations area operations, including tritium operations. As expected, the early years 

(1950s/1960s) have substantially fewer entries compared to later years. Statistics of the number 

of entries by year, facility code, and area code were compiled into a Microsoft Excel spreadsheet 

and are shown in Attachment 2 to this report. Several generic examples of incidents were also 

compiled in a spreadsheet to inform the Work Group more effectively of the types of incidents 

related to radiological exposure. Attachment 3 to this report provides the recommended 

selection criteria for the retrieval of pertinent incidents in the WSMS database. 

The WSMS incident database contains thousands of entries relating to worker uptakes, high 

airborne concentrations, wounds, high dose rate exposure situations, skin contaminations, and 

bioassays. These data are relevant to the identification of incidents by dose reconstructors and 

provide valuable information for the development and/or verification of efficiency methods. In 

order to link the incidents in this file with individual workers, it is essential to identify and 

review the source data. Some of these source documents are readily available to NIOSH. 

However, the database also contains incidents that are not of interest in dose reconstruction, such 

as basketball, skiing, and false fire alarms. Therefore, information from this database should be 

requested based on the information compiled in the spreadsheet of generic examples and 

available documentation to isolate radiological incidents. Since it does not adequately represent 

all areas of the site, the WSMS incident database cannot be used as the sole source of SRS 

incidents. However, a combination of the incident information in the database and the source 

document information would provide one mechanism for determining the completeness of 

incidents in the individual Health Physics file. The entries in the WSMS incident database were 

compared to those abstracted from the tank farm Fault Tree Data Bank in Makhijani et al. 

(1986), as previously discussed in Comment 4. 

Most of the entries reviewed from the tank farm Fault Tree Data Bank were not located in the 

WSMS incident database. As previously mentioned, this may reflect a difference in operational 

scope between the two collections of incident data. SRS requested Central Records to search the 

records database for “Fault Tree.” However, the observations made to this point imply that 

multiple Fault Tree Data Banks have been maintained, corresponding to the various areas of the 

site. For example, the raw materials area has its own data bank. 

Additional data may also be available for this purpose for the period 1956 to 1965 in Works 

Technical Department progress reports (DPSP). These reports provide insights into facilityspecific high-risk job exposures relative to maintenance and repair of reactors; entries into very 

high dose rate areas in the F- and H-Canyons; repair and maintenance of contaminated 

equipment, suck-backs, and outside ambient exposures from process buildings; and waste tank 

maintenance, repair and surveillance of high-level waste tanks. During this period, SRS 

management clearly recognized that these activities posed potentially high risks. 

The database also refers the reader back to the source document, which may contain additional 

information on the incident. Having incident information available will also lift the burden of 

providing this information during the CATIs and bring some equality to this process. 

The data review described above appears to underscore the validity of this concern. Multiple 

databases from various areas of SRS contain a wealth of data regarding incidents and exposures 

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throughout the site’s history, and it is clear that all incidents are not reflected in worker exposure 

records or other radiological records. In order to ensure consistently claimant-favorable dose 

reconstructions, NIOSH should ensure that incidents involving significant internal or external 

exposure are either captured in worker exposure records or made accessible to dose 

reconstructors, with appropriate guidance for when to use such records. 

Without a thorough reconciliation of the DOE exposure files against these separate incident data 

banks, NIOSH cannot be assured that all significant exposures from incidents are considered in 

relation to individual worker claims or to high-five estimates of maximal doses. These tools 

should be made available to dose reconstructors as an additional reference. 

High-Risk Exposure 

In its previous review, SC&A recommended including guidance within the TBD on how to 

address special exposure conditions. SC&A listed several groups of workers (e.g., construction, 

subcontract, decontamination/decommissioning) and authorized or unauthorized practices 

(e.g., recovering/processing U-233 and thorium, producing transplutonium radionuclides, 

burning tributyl phosphate, opening high-level waste tank risers, and eating contaminated 

foodstuffs) that may have experienced or caused significant unrecognized or unreported 

exposures (SC&A 2005a). 

Additional data may also be available related to high-risk job exposures in Works Technical 

Department progress reports (DPSP). These reports provide insights relative to maintenance and 

repair of reactors; entries into very high dose rate areas in the F- and H-Canyons; repair and 

maintenance of contaminated equipment, suck-backs, and outside ambient exposures from 

process buildings; and waste tank maintenance, repair, and surveillance of high-level waste 

tanks. SRS management clearly recognized that these activities posed potentially high risks. 

For consistency among dose reconstructions, the reviewers concluded that the TBD should alert 

the dose reconstructor to special conditions when a deviation from the standard dose 

reconstruction methodology is needed. 

4.13.2 Applicability to Revision 03 

The lack of consideration of incidents and high-risk jobs continues to be applicable to 

Revision 03. Some progress has been made towards resolution in this area; however, this 

occurred after the release of Revision 03. 

4.13.3 Summary of Overall NIOSH Response 

The dose of record information already contains significant dose to workers from 

incidents. In many cases, the workers will state that no incident occurred, but the records 

do, in fact, identify incidents. It is possible for a claimant to say that an incident occurred 

but the records do not identify any such incident. Not many cases exist in which a 

worker states that an incident and the dose record does not include it. The high-five 

approach is used as a means to ensure that missing an incident during the performance of 

a dose reconstruction will not result in an underestimate of the reconstructed doses. In 

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addition, Form OCAS-INT-004 requests DOE to submit relevant incident information for 

consideration in dose reconstruction. 

NIOSH provided information in the OCAS response to SC&A on June 6, 2006, and via 

comments during the Work Group call on August 22, 2006, that it obtained from staff at 

SRS, who told NIOSH that the WSMS incident database is no longer maintained at SRS, 

and that it is maintained by a contractor, who would provide a redacted database for a 

charge. NIOSH has since contacted both WSMS as owner of the database and SRS as 

owner of the content. NIOSH spoke with the Nuclear Safety Manager at SRS, who 

promised to work with the DOE point of contact at SRS to work out details for providing 

the database to ORAUT, through the assistance of WSMS. 

Work Group Actions 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Request of copy of the SHI reports from DOE/WSRC in writing. 

•	 Obtain the user’s guide for the WSMS database from DOE/WSRC. 

•	 Research the pedigree of the WSMS database reviewed during the February 28–March 1, 

2007, visit to SRS and determine the source database for Makhijani et al. (1986). 

a.	 Follow up to determine if there is a record of the Freedom of Information Act 

request associated with Makhijani et al. (1986). 

b.	 Follow up with the original author of the database reviewed during the site visit. 

•	 Check on the availability of other incident databases and reports (e.g., Waste 

Management Incident, Raw Materials Incident, Tank Farms Incident, Separations 

Incident, Reactor Incident) 

SC&A Action Items: 

•	 Have the excess spreadsheet (queries) reviewed for release by DOE and forward them to 

the Work Group, NIOSH, and ORAUT. (Completed) 

•	 In conjunction with the Work Group, provide queries isolating records of interest in the 

WSMS database. (Completed; see Attachment 3 to this report) 

4.13.4 Closure Status 

SC&A maintains that some workers have sustained significant exposures that are relevant to 

their dose reconstruction but may not be captured by reviewing the DOE exposure file and the 

CATI. Sources other than dosimetry files exist at SRS that summarize incidents and provide 

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supporting documentation. Databases identified may be specific to an area or operation, so a 

compilation of several sources may be necessary to capture sitewide incidents. The databases 

vary in the criteria for inclusion, giving a broader perspective on what constitutes an incident. 

Without a thorough reconciliation of the DOE exposure files against these separate incident data 

banks, NIOSH cannot be assured that all significant exposures from incidents are considered in 

relation to individual worker claims or in efficiency methods (e.g., high-five approach, coworker 

dose assignments). Dose reconstructors should be alerted to these situations. Based on record 

storage practices, redundant systems are necessary to develop a complete list of incidents. While 

the CATI provides some potential to identify incidents that may be missing from an individual’s 

dose record, reliance on the CATI for this purpose places an inappropriate burden upon the 

worker or survivor to recall events and to recognize potential implications for the dose 

reconstruction. 

Revision 03 gave no additional consideration to many potentially unrecognized or unreported 

exposures from high-risk work activities. Revision 03 did include a discussion of dose from the 

ingestion of foodstuffs (fruits and vegetables) containing radioactive material. Proposed 

Revision 04-E (Scalsky 2006) has discussions on potential dose from special campaigns 

involving U-233, thorium, and transplutonium radionuclides. ORAUT-OTIB-0052 (Chew et al. 

2006) addresses dose to construction workers. Neither Revision 03 (Scalsky 2005) or the 

proposed Revision 04-E (Scalsky 2006) of the TBD include information on decontamination and 

decommissioning, open burning of solvents, and other high-risk or unusual jobs. It is 

recommended that this issue remains open pending the completion of Work Group actions and 

the release of proposed Revision 04-E (Scalsky 2006) of the site profile. 

4.14	 COMMENT 14: AVAILABILITY OF ADDITIONAL SOURCE DOCUMENTS 

AND DATA 

Based on searches performed for records pertaining to dose reconstruction, SC&A has identified 

additional sources of dosimetry data, primarily neutron dosimetry data that are not currently 

provided to NIOSH/ORAUT by SRS, but that may be important to dose reconstruction. 

4.14.1 Issue Description 

4.14.1.1 Neutron Data 

SC&A pointed out in the original review that logbooks containing neutron dose data exist but the 

information was not being provided to NIOSH in the claimant dosimetry file because SRS had 

not retrieved and scanned these logbooks (as of February 28, 2007) (Robertson-DeMers 2007). 

As previously stated, the most complete records at SRS from a population standpoint are the 

hardcopy records. Prior to the last Work Group meeting, SC&A asked the Radiological Records 

Manager at SRS the following question: 

Have you retrieved the neutron exposure logbooks for 1963–1972? If so, is this 

data provided to NIOSH when requesting claimant information? 

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He responded as follows: 

No. We do not have time allocated for this search. 

Specifically, for 1963–1972, three sets of logbooks contain data for composite (beta/gamma + 

tritium + neutron) doses, tritium dose only, and neutron dose only. These logbooks would be 

necessary to decompose the composite data to perform best estimate dose reconstruction. The 

availability of NTA data not provided in the claimant files was reinforced by the Radiological 

Records Manager during a meeting with Sam Glover (NIOSH), Elyse Thomas (ORAUT), Mark 

Griffin (Advisory Board), Brad Clawson (Advisory Board), and Kathryn Robertson-DeMers 

(SC&A). 

SC&A provided a list of pertinent dosimetry records to NIOSH for consideration. In addition to 

the data sources identified in the site profile, History Associated Incorporated prepared sitespecific guides to epidemiologic and health-related records at a number of sites including SRS. 

A list of records available at SRS is available through the DOE Office of Epidemiologic Studies 

Web site. This inventory of records includes the following (DOE 1995): 

•	 Densitometry Records (Personnel Monitoring Film Badge Data), 1952–1976, 

1979, 1981–1984, Health Physics Section and Health Protection Department, E.I. 

du Pont de Nemours and Company.... 

Series Description: This record series contains personnel monitoring data for 

employees and visitors at the Savannah River Site. Data sheets and computergenerated reports provide exposure data collected from film badges, neutron 

pencils, and range ring dosimeters. Information includes employee name, payroll 

number, dosimeter number, health protection area, cycle or pull date, and 

readings. X-ray and gamma exposures, slow neutron exposures, and current and 

cumulative open window and shield dosimeter readings are given, usually in 

millirem. Calibration data includes densitometer, emulsion, code number, 

developer and reader name, developing temperature, and beta and gamma results 

in millirem. 

•	 Neutron Exposure Reports, 1971, Health Physics Section, E.I. du Pont de 

Nemours and Company.... 

Series Description: This series consists of neutron exposure reports and related 

correspondence. Information includes employee name, payroll number, 

thermoluminescent badge number, work area, cycle number, date badge was 

issued and returned, code number, and dose in millirems. 

•	 Dosimetry Logs, 1954–1978, Health Physics Section, E.I. du Pont de Nemours 

and Company... 

Series Description: This series consists of logs that document the distribution of 

dosimeters and associated results. Dosimeters include finger rings, film badges, 

thermoluminescent dosimeters, thermoluminescent neutron dosimeters, and 

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“NTA” badges. Information includes frequency of dosimeter changes and 

readings, number distributed to certain areas, calibration data, and results for 

individual employees. 

•	 Neutron Dosimetry Data Logs, 1972–1982, 1986–1992, Health Physics Section, 

E.I. du Pont de Nemours and Company... 

Series Description: This series consists of logs which record the assignment of 

thermoluminescent dosimeter (TLD) neutron badges to Savannah River Site 

employees. Information includes TLD badge number, name, social security or 

payroll number, supervisor's name and telephone number, date annealed and 

processed, and cycle number. There are separate sheets for temporary badges. 

Dates and locations are listed on each sheet. Readings are usually given in 

millirems. The series also includes Neutron Badges Indices listing the total 

number of badges issued, extra issues, badges read for cycle, controls, late 

badges, badges not used, and badges not returned. 

•	 Neutron Pencil Results (1952), Health Physics Section, E.I. du Pont de Nemours 

and Company... 

Series Description: This series consists of lists recording neutron pencil 

dosimeter results. Information includes employee name, payroll number, neutron 

dose in millirems, pencil identification number, and reason why doses could not 

be determined from some pencils. 

•	 NTA Film Badge Inventory and Calibration Sheets, (1950–1962, 1964–1969)... 

Series Description: This series consists of “NTA” film badge inventory and 

calibration sheets and memoranda regarding the “NTA” monitoring program. 

Information concerns the collection and processing of “NTA” film badges, 

procedures, suggested revisions to the program, areas where badges were 

collected, and the date received. “NTA” badge and payroll numbers are 

included. Some records include employee names, as well as other personnel 

information. 

A review of several records storage transfer request forms from the Health Physics Section was 

conducted by Kathryn Robertson-DeMers (SC&A). Documents identified as containing neutron 

exposure information for the period of 1963–1972 are listed below: 

•	 Transmittal dated December 29, 1971: 

NTA Film, 1962 through Cycle 8, 1967, Document #T2092 

NTA Film, Cycle 9, 1967 thru Cycle 6A, 1970, Document #T2091 

NTA Film, Cycle 6B, 1970 thru Cycle 9B 1971, Document #T2101 

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•	 Transmittal dated March 14, 1973: 

Neutron Data 1970 and 1971, Document #M440 

The combination of sources above should provide a complete history for neutron dose from 

1963–1972. NIOSH may want to consider performing a search of the site records database, 

especially as it pertains to information related to neutron exposure data for individuals. This 

additional source of data may be relevant to the assignment of neutron dose. 

The Web site also lists other pertinent documents related to the SRS evaluation: 

•	 Film Processing and Dosimeter Calibration Procedures (1954–1960) 

•	 Daily Badge Processing Reports (1954–1957) 

•	 Employee Radiation Exposure Record Cards (1964–1992) 

•	 Construction Worker Exposures (1958–1959) 

•	 Bioassay Analysis Reports (1979, 1981, 1989–1992) 

•	 Bioassay and Dosimeter Data Sheets (1961–1972, 1975–1977, 1986–1988) 

•	 Bioassay Logs (1953–1960, 1962–1989, 1991–1992, 1995) 

•	 Bioassay Master Report (1975) 

•	 Bioassay Monthly Reports (1982) 

•	 Bioassay Program Records (1953–1954, 1992) 

•	 Bioassay Results—Fission Products Induced Activity Reports (1965, 1967–1977, 1987– 

1989) 

•	 Bioassay Sampling Cards (1954–1987) 

•	 Contamination Cases Logbooks (ca. 1977–1992) 

•	 Contamination Incident Reports (1951–1955, 1958, 1971–Current) 

•	 In Vivo Count Results (1982–1983, 1985) 

•	 Tritium Doses (1958–1972) 

•	 Tritium Dosimetry Monthly Reports (1972–1976, 1979, 1981–1989, 1992–1993) 

•	 Tritium Dosimetry Reports (1977–1979, 1983–1987, 1992–1993) 

•	 Type A Occurrence Report (1979) 

•	 Type C Occurrence Reports (1972–1981) 

•	 Uranium Concentration Reports (1989–1990) 

•	 Urinalysis Reports Strontium and Plutonium (1984–1990) 

These records may assist NIOSH in its evaluation of the high-five approach, tritium, and 

incidents. 

Completeness of HPAREH 

Another issue relates to the fact that the HPAREH database does not contain data for all the 

workers for 1952 onward. It contains data for those that were monitored beginning in 1979 and 

afterwards. It also contains data for some (but not most) workers who worked during 1952– 

1978. This is important to realize when constructing a coworker model for assigning doses, such 

as OTIB-0032 (Merwin 2006) for beta/gamma doses and if a neutron coworker dose model is 

constructed in the future. The dose data from the earlier years (1952–1978) for a large 

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population of workers should be included to create realistic year-by-year dose assignment tables 

for unmonitored workers. It is especially important to include these early years of dose data 

because doses during these early years most likely were greater than later years as radiation 

protection practices/knowledge improved. 

NIOSH indicates that it presented the issue of what records from prior to 1979 were included in 

the HPAREH database to an SRS staff member, who replied in an email dated August 24, 2006. 

NIOSH has not provided the specific questions asked, making it difficult to understand the 

answer. An SRS staff member did, however, state the following in his response to NIOSH: 

The latest version of HPAREH at SRS includes the complete dose history for more 

than 70% of the individuals ever monitored at SRS, and more than 70% of the 

cumulative collective worker doses delivered at SRS. It is important to note that 

SRS staff has never envisioned (or used) HPAREH as the only component of the 

official dose of record for any worker at SRS. The official dose of record includes 

all the forms of dose records available to us. 

As previously stated above, SRS believes that the most complete records at SRS from a 

population standpoint are the hardcopy records. Comment 5 above provides a detailed 

discussion of the limitations of HPAREH. 

Multiple Dosimetry Data 

A number of conditions can result in partial body exposures or portions of the body being 

exposed unevenly. SRS recognized this early in its operations and implemented a multiple 

dosimetry program. Multiple dosimeters were required where measurements showed 

nonuniform radiation fields (e.g., specific areas within the reactors required head monitoring, as 

the dose to the head was greater than that to the chest). The term “multiple badging” in this 

context refers to instances where workers wore more than one badge at the same time in order to 

capture doses to various parts of the body at risk of greater exposure more accurately than would 

be indicated by a single dosimeter worn at the pocket level. The TBD has not mentioned the 

multiple badging programs and how this information could be applied to assigning organ dose. 

During site expert interviews, SC&A learned that multiple badge results are not routinely 

included in the personnel radiation exposure record and are not currently provided to 

NIOSH/ORAUT. The whole-body dose for an individual wearing multiple badges was assumed 

to be the highest recorded result on any of the badges. For example, if the results from four 

badges positioned over the body were 100 mrem, 200 mrem, 300 mrem and 400 mrem, the 

whole-body dose would be recorded as 400 mrem regardless of the position of the badge. In 

about 1992, the methodology for assigning whole-body dose from multiple badges changed. 

Each dosimeter was assigned an effective whole-body dose equivalent, and the resulting values 

were added to obtain the whole-body dose. In this case, weighting factors from ICRP 26, 

Recommendations of the International on Radiation Protection, are used (ICRP 1977). The 

whole-body dose is the only number that is documented in the individual’s dosimetry file. All 

dosimeters worn were processed and results were recorded by area of the body. 

NIOSH/ORAUT should evaluate the results of multiple dosimeter processing by body position to 

determine whether partial-body exposures are an issue for specific organs and to evaluate 

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whether the dose values provided by the multiple dosimetry are more claimant favorable. As 

individual files do not routinely include this information, a special request will be required. 

4.14.2 Applicability to Revision 03 

This comment is applicable to Revision 03 of the SRS TBD. Draft Revision 04-E (Scalsky 

2006) provides no additional information to resolve this issue. 

4.14.3 Summary of Overall NIOSH Response 

Initially, NIOSH stated that the nature of this comment was unclear. It referred SC&A to the 

additional guidance in OCAS-TIB-006, Interpretation of External Dosimetry Records at the 

Savannah River Site (SRS) (Neton 2004), and OCAS-TIB-007, Neutron Exposures at the 

Savannah River Site (Neton 2003c), prepared following preparation of Section 5.0 in the SRS 

site profile in 2003. NIOSH examined the original SC&A review for clarification regarding 

logbooks. 

NIOSH issued OCAS-PER-019, The Effect of Additional Neutron Data from the Savannah River 

Site (OCAS 2007). This program evaluation report indicated that SRS provided additional 

neutron data. NIOSH reevaluated 17 claims, including four noncompensable claims. No 

changes occurred in the POC. Previously unevaluated claims were completed with the new data. 

4.14.4 Work Group Action 

The following actions were identified in Work Group deliberations: 

NIOSH Actions: 

•	 Review neutron logbooks referenced in the August 22, 2006, meeting. The meeting 

minutes are available on the NIOSH Web site. 

•	 Provide a response to the Board regarding the completeness of the HPAREH file used for 

the development of the external coworker model. 

•	 Investigate the sources of information provided by SC&A. 

4.14.5 Closure Status 

Revisions 03 and draft Revision 04-E (Scalsky 2006) of the TBD lack an explanation of why the 

additional records are not being considered for dose reconstruction. SC&A provided NIOSH 

with an inventory of supplemental records that may be beneficial in dose reconstruction. Based 

on a review of OCAS-PER-019 (Allen 2007), SRS provided some additional neutron data. 

Whether these are the neutron logbook data mentioned above is unknown. To date, NIOSH has 

not discussed the exact nature of these data. NIOSH has made no effort to evaluate the 

completeness of the HPAREH file used in the development of the external coworker model. The 

integrity of the HPAREH file for use in coworker modeling is questionable given the absence of 

data for most workers terminating prior to 1979. Either the TBD or the external coworker dose 

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procedure should develop and discuss a basis for its appropriateness. Furthermore, consideration 

should be given to multiple dosimetry results, which may provide valuable information on 

nonuniform exposures and may be of assistance in determining organ dose for some workers. 

For the above reasons, it is recommended that this overall finding remains open pending review 

of the NIOSH Work Group action items. 

4.15 COMMENT 15: AMBIGUOUS DOSE RECONSTRUCTION DIRECTION 

Well-developed technical documents are necessary to ensure that methods are effective and 

consistent. Applying consistent methodologies can provide continuity of assessment over time 

and across multiple facilities. The SRS TBD contains ambiguous instructions, inconsistencies, 

and unwarranted precision. 

4.15.1 Issue Description 

Dose reconstruction is a complex process even under the best circumstances. It is, therefore, 

imperative that supportive background information/data and specific instructions are presented in 

a logical manner that ensures understanding, process efficiency, and consistency among dose 

reconstructors. Many of the sections of the TBD, especially Chapter 4 related to internal 

dosimetry, are very difficult to understand and, together with the large array of TIBs and other 

OCAS/ORAUT procedures, create a virtually impenetrable, complex array of guidelines. This 

situation lends itself to inconsistencies in the way in which dose reconstructions are performed 

and makes it difficult to verify the reliability and reproducibility of the dose reconstructions. A 

major factor that limits the readability and, therefore, comprehensibility of Section 5.0 of the 

TBD is the mingling presentation of data that alternates between beta/photon and neutron 

dosimeters/dosimetry. Since reconstruction of beta/photon and neutron exposures requires two 

different methods, as well as IREP inputs, a more logical and comprehensible format would have 

separated these two major topics. 

Quality assurance is an important part of maintaining a consistent and defensible dose 

reconstruction program. NIOSH/ORAUT should make the TBD transparent to the user and 

ensure that the various portions of the TBD are consistent with one another. Inconsistencies in 

the TBD and between the TBD and other procedures result in confusion and a potential 

misapplication of available dose reconstruction methods, and they should be corrected or 

explained. The issuance of complexwide TIBs designed to address specific issues such as 

glovebox work, tritium exposure, ambient environmental dose, and medical x-ray exposure 

among others has provided clarification. Procedures are used in conjunction with TIBs and site 

profiles to guide the dose reconstruction process. This provides more consistency in the process 

overall. 

4.15.2 Applicability to Revision 03 

These comments are applicable to Revision 03 of the TBD. 

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4.15.3 Summary of Overall NIOSH Response 

A large number of TIBs have indeed been developed to address unique situations or very 

general, complexwide situations, all designed to address issues not easily covered in site profiles 

or to add efficiency to the POC evaluation process. Procedures are written, in part, to help 

ensure that the dose reconstructors consider all the various guidance documents. Site profiles, 

TIBs, and procedures are all essential aids to assist the dose reconstructors in performing their 

POC evaluations. Indeed, the lack of these documents would likely lead to less consistency and 

efficiency. 

4.15.4 Work Group Action 

There are no Work Group actions pending. 

4.15.5 Closure Status 

This issue is currently being addressed under a separate review task. The additions and updates 

to TIBs and procedures have clarified some of the ambiguity in the site profile. The use of site 

workbooks by dose reconstructors ensures that methods are consistent. The mingling of data 

causes confusion in the identification of dose reconstruction methodologies. Some improvement 

has been made in this area with the later revisions of the TBD. It is recommended that this issue 

be considered closed. 

4.16	 COMMENT 16: SPECIAL EXPOSURE CIRCUMSTANCES FOR 

SUBCONTRACTORS AND CONSTRUCTION WORKERS 

The TBD does not currently include the special exposure circumstances for subcontractors and 

construction workers; however, NIOSH is aware of this issue. ORAUT-OTIB-0052 (Chew et al. 

2006) was developed to provide dose reconstruction guidance for trade workers. SC&A is 

currently reviewing this procedure under Task 3. 

4.16.1 Applicability to Revision 03 

The trade worker section in Revision 03 of the TBD is reserved. NIOSH issued a generic TIB, 

ORAUT-OTIB-0052 (Chew et al. 2006), providing dose estimate parameters for construction 

workers, which includes consideration of SRS construction workers. 

4.16.2 Summary of Overall NIOSH Response 

No response was provided. 

4.16.3 Work Group Action 

This is no longer an issue and there is no Work Group action pending; therefore this issue is 

closed. 

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4.16.4 Summary 

The issuance of ORAUT-OTIB-0052 (Chew et al. 2006) provides a mechanism for estimating 

construction worker doses. It has been included in the Task 3 procedure reviews and will not be 

discussed in this review. It is recommended that with the issuance of the construction worker 

procedure, this issue be considered closed and requires no further action. SC&A recommends 

that NIOSH finish the trade worker section in the SRS TBD for completeness. 

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5.0	 OVERALL ADEQUACY OF THE SRS SITE PROFILE, 

REVISION 03, AS A BASIS FOR DOSE RECONSTRUCTION 

The SC&A procedures call for both a “vertical” assessment of a site profile in terms of the 

adequacy and completeness of each particular element of the profile as well as a “horizontal” 

assessment of the extent to which the profile as a whole satisfies its intended purpose and scope. 

This section addresses the latter objective by evaluating (1) how, and to what extent, the site 

profile satisfies the five objectives defined by the Advisory Board for determining adequacy; 

(2) the usability of the site profile for its intended purpose (i.e., to provide a generalized technical 

resource for the dose reconstructor when individual dose records are unavailable), and 

(3) generic technical or policy issues that transcend any single site profile that need to be 

addressed by the Advisory Board and NIOSH. 

5.1 SITE PROFILE IMPROVEMENTS 

In general, Revision 03 of the TBD included minimal changes, including an analysis of the 

maximum plausible annual dose from the ingestion of foodstuffs at SRS. This issue was 

originally raised in a worker outreach meeting and integrated to respond to this comment. Draft 

Revision 04-E (Scalsky 2006) proposes additional improvements. 

5.2 SATISFYING THE FIVE OBJECTIVES 

The SC&A review procedures, as approved by the Advisory Board, require that each site profile 

be evaluated against five measures of adequacy—completeness of data sources, technical 

accuracy, adequacy of data, site profile consistency, and regulatory compliance. The SC&A 

finds that Revision 03 of the SRS site profile represents an adequate accounting of the primary 

internal issues related to plutonium, uranium, and tritium, as well as main external hazards. 

Revision 03 is not substantially different than Revision 02 and therefore only minimally resolved 

the 16 issues identified in the Revision 02 review. The primary changes made to Revision 03 

included incorporation of internal review comments and inclusion of a model for the ingestion of 

foodstuffs. Therefore, Revision 03 of the SRS site profile falls short in fully characterizing a 

number of key underlying issues that are fundamental to guiding dose reconstruction. In some 

cases, these issues may impact other site profiles. Many of the issues involve a lack of sufficient 

conservatism in key assumptions or estimation approaches, or incomplete site data or analyses of 

these data. 

SC&A is aware of NIOSH’s ongoing efforts to develop and issue Revision 04-E (Scalsky 2006) 

of the SRS site profile and has acknowledged additional information intended for inclusion in 

that version. Likewise, additional guidance documents were being issued that, while not yet 

reflected in Revision 03 of the SRS site profile, would serve to mitigate some of the gaps and 

issues raised in this report. While Revision 04-E (Scalsky 2006) incorporates changes that will 

close a number of the comments, it remains a draft and has not been formally released. 

Therefore, this report continues to cite issues related to those intended improvements as “open” 

pending issuance of Revision 04-E (Scalsky 2006). 

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5.2.1 Objective 1: Completeness of Data Sources 

Revision 03 of the TBD contains incomplete assessment and guidance on dose assignment 

pertaining to RU, Pu-242, transplutonium radionuclides, thorium, and U-233. The calculation of 

internal dose does not explicitly consider transuranic and fission product impurities in RU. 

Impurity concentrations are based on estimates made from waste stream data and appear to be at 

odds with the DOE 1985 Task Force review on RU (DOE 1985). A further investigation of the 

RU source term data should be completed to determine the upper bounds of impurity 

concentrations and resulting doses. Other assays, such as metallurgical analyses, may assist in 

determining concentrations and relative uncertainties in these values. Revision 03 does not 

address potential internal dose from U-233 (including impurities in uranium). Revision 04-E 

(Scalsky 2006) adds discussions on the assignment of dose from impurities in RU and from 

U-233; however, the document has not been formally released. 

Revision 03 does not consider potential contributions from exposure plutonium containing 

higher levels of Pu-242, nor does it justify the absence of such a discussion. Proposed 

Revision 04-E (Scalsky 2006) discusses the production of Pu-242 during curium campaigns and 

provides information on the activity composition of the high Pu-242 mixture. The default 

assumption for plutonium remains at 10-year-old 12% plutonium. 

SRS handled exotic radionuclides as a result of special production campaigns and source 

production ranging in quantities from fractions of a gram to kilograms during special campaigns. 

These radionuclides included transplutonium elements, Po-210, Co-60, Cf-252, Tm-170, Ir-192, 

Eu-152, and various isotopes of lanthanum (Reed et al. 2002). Many of these sources produced 

were encapsulated and therefore posed primarily an external hazard. Neptunium and curium 

were also processed for periods of time. Revision 03 of the TBD does not explicitly cover the 

potential dose from these radionuclides. The document gives inadequate or no consideration to 

potential exposures and missed dose from these radionuclides and does not discuss the 

implementation of monitoring techniques for these radionuclides. 

Section 4.1.2 on bioassay was added in Revision 03, but the issues related to RU, Pu-242, 

transplutonium radionuclides, thorium and U-233 have not been adequately resolved. Several 

updates proposed for Revision 04-E (Scalsky 2006) of the TBD deal directly with exposure from 

these radionuclide sources. This will bridge some of the gaps found in previous versions of the 

TBD. It is recommended that issues related to exposure from RU, Pu-242, transplutonium 

radionuclides, U-233 and other exotic radionuclides remain open pending review of the yet-tobe-released Revision 04-E (Scalsky 2006) of the TBD and ORAUT-OTIB-0053 (ORAUT 2005). 

SC&A maintains that some workers have sustained significant exposures that are relevant to 

their dose reconstruction but may not be captured by relying on the DOE exposure file and the 

CATI. Other sources besides dosimetry files exist at SRS that summarize incidents and provide 

supporting documentation. Databases identified may be specific to an area or operation, so a 

compilation of several sources may be necessary to capture site-wide incidents. The databases 

vary in the criteria for inclusion, giving a broader perspective on what constitutes an incident. 

Without a thorough reconciliation of the DOE exposure files against these separate incident data 

banks, NIOSH cannot be assured that all significant exposures from incidents are considered in 

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relation to individual worker claims or in efficiency methods (e.g., high-five approach, coworker 

dose assignments). Dose reconstructors should be alerted to these situations. Based on record 

storage practices, redundant systems are necessary to develop a complete list of incidents. While 

the CATI provides some potential to identify incidents that may be missing from an individual’s 

dose record, reliance on the CATI for this purpose places an inappropriate burden upon the 

worker or survivor to recall events and to recognize potential implications for the dose 

reconstruction. 

Revision 03 gave no additional consideration to many potential unrecognized or unreported 

exposures from high-risk work activities. Proposed Revision 04-E (Scalsky 2006) includes 

discussions on potential dose from special campaigns involving U-233, thorium, and 

transplutonium radionuclides, and ORAUT-OTIB-0052 (Chew et al. 2006) addresses dose to 

construction workers. However, neither Revision 03 nor proposed Revision 04-E (Scalsky 2006) 

of the TBD includes information on decontamination and decommissioning, open burning of 

solvents, and other high-risk or unusual jobs. It is recommended that this issue remains open 

pending the completion of Work Group actions and release of proposed Revision 04-E (Scalsky 

2006) of the site profile. 

The completeness of the data used as a basis for the high-five approach is questionable. NIOSH 

has not reviewed all recorded inhalation intakes over the history of the site. The criteria for 

inclusion of individuals in the SRS intake file have changed over time, excluding some intakes 

from the file. Other sources of information such as the DPSP monthly reports, databanks and 

incident files, and visitor cards were not considered when identifying intakes. SC&A has 

presented numerous examples from these sources that likely meet the criteria for inclusion in the 

high-five approach and deserve further investigation. The general bioassay trend for Np-237 and 

Pu-239 from the individual high-five data and the DPSP monthly report data indicate that 

monthly report bioassay results were higher. In addition, some radionuclides were present at the 

site prior to the availability of bioassay techniques. It is recommended that several Work Group 

actions remain open, including clarification of the location of the SRS high-five bioassay data, 

review of the tank farm Fault Tree Data Bank, and completion of the revised high-five approach. 

The issuance of ORAUT-OTIB-0052 (Chew et al. 2006) provides a mechanism for the 

estimation of construction worker doses. It has been included in the Task 3 procedure reviews 

and will not be discussed in this review. It is recommended that with the issuance of the 

construction worker procedure, this issue be considered closed and requires no further action. 

SC&A recommends that NIOSH complete the trade worker section in the SRS TBD for 

completeness. 

5.2.2 Objective 2: Technical Accuracy 

The method used to reconstruct doses to unmonitored outdoor workers due to airborne emissions 

employs an atmospheric dispersion model, assumptions, and a resuspension factor that do not 

appear to be claimant favorable and are not entirely appropriate for this class of problem. 

Revision 03 (Scalsky 2005) and proposed Revision 04-E (Scalsky 2006) provide no clarification 

on these issues. In its response to SC&A’s finding, NIOSH maintains that the dispersion model 

has been adequately validated and is sufficiently conservative. It has not specifically addressed 

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ground-level plumes from open-pan burning of contaminated solvents or from environmental 

spills and leaks in the TBD. Furthermore, it has not provided a written evaluation of the existing 

dispersion model focusing on episodic releases, as proposed by the Work Group. 

The TBD lacks a basis for the use of a resuspension factor of 1×10-9 for resuspension of 

contaminated soil. Based on an SC&A review of the literature, it also appears that the TBD 

resuspension factor of 1×10-9 per meter may not be claimant favorable. Kennedy and Strenge 

(1992) reported resuspension factors from approximately 1×10-11 to 1×10-2

 m

-1, which suggests 

that resuspension is a complex process of several parameters and that the specific conditions 

present at the time of measurement are critical. Based on recommended resuspension factors 

presented in the literature, an average value closer to 1×10-5 to 1×10-6 per meter would seem 

more appropriate for use in worker dose reconstruction, resulting in worker inhalation doses 

from resuspension that are 3 to 4 orders of magnitude greater than those derived in the site 

profile. The dust-loading approach should also be considered, using an average work-year dust 

loading on the order of perhaps 1 mg/m3 . It is recommended that, due to remaining issues 

associated with the atmospheric dispersion model, assumptions, and a resuspension factor, this 

overall finding remains open. 

Revision 03 contains inadequate information regarding the assessment of dose from STCs. In 

June 2006, NIOSH proposed a methodology for the assignment of dose from STCs. Proposed 

SRS-specific guidance assigns dose from tritides based on surface contamination limits rather 

than production information and surveillance data, making the basis for assumptions weak, 

particularly in years when engineering controls were not as advanced as they are today. In April 

2007, NIOSH released ORAUT-OTIB-0066 (LaBone 2007), which provides a bounding 

technique for the assignment of dose from intakes of OBTs and SMTs. NIOSH has provided a 

dose estimation methodology for STCs; however, it has not verified the timeframe and location 

where STCs were handled or the types and quantities of STCs handled at SRS. The bounding 

techniques, as proposed for SRS, cannot be effectively developed and applied without some 

basic understanding of the STCs handled, the quantities of material, the locations and time 

periods of potential exposure, and the physical behaviors of tritium compounds in the 

environment (e.g., conversion to HTO, formation of rust) to correctly characterize tritium 

exposure. Furthermore, NIOSH limited the application of this technique to 1975 to present. 

This conflicts with site expert statements that indicate that potential exposures occurred as far 

back as late 1950s. Given the large amount of tritium handled at SRS in various areas and the 

propensity of tritium to bind with organics and metals, it is reasonable to assume that STCs were 

present at SRS in some form prior to 1975. An evaluation of the adequacy of the dose estimation 

methodology cannot be completed without this key information; thus it is recommended that the 

issue remains open. 

In-vivo bioassay monitoring results for thorium should be analyzed very carefully. Errors of two 

orders of magnitude can be made, depending on the material type, equilibrium assumptions, and 

time of measurement after intake. Urinalysis results should also require a very cautious analysis, 

including the influence of natural thorium in the diet. 

Information in the revised TBD refers the dose reconstructor to ORAUT-OTIB-0001 (Brackett 

2003), which describes the high-five approach. NIOSH indicated that it intends to update the 

high-five approach and base revised calculations on bioassay data rather than data in the SRS 

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IDR. This update was expected in December 2006 but has not been completed as of August 

2007. Until the revision of this procedure is implemented and the TBD references the correct 

TIB, regulatory issues continue to be associated with the application of the high-five approach. 

Using the urinalysis as a basis for dose calculation will eliminate the use of intake data derived 

with ICRP 30 and allow for the exclusive use of ICRP 60 methodology or more current methods. 

Revision 03 (Scalsky 2005) and Revision 04-E (Scalsky 2006) still reference the 2003 version of 

ORAUT-OTIB-0001 (Brackett 2003) on maximizing internal dose. The updated approach along 

with any modifications to the TIB will be reviewed when provided by NIOSH. 

Revisions 03 and 04-E reference ORAUT-OTIB-0001 (Brackett 2003), which continues to use 

both ICRP 30 and subsequently ICRP 66 as the method for assignment of internal dose. 

ORAUT-OTIB-0001 (Brackett 2003) justifies the use of intakes calculated with ICRP 30 

methodologies rather than the most current ICRP methodology by comparing IRFs from ICRP 

30 and ICRP 68. The ICRP 30 model does not produce intake values that are higher than those 

derived by the new ICRP models for a majority of the relevant radionuclides included in the 

hypothetical intake as maintained by NIOSH. The use of ICRP 30 methodology to calculate the 

intake, with a subsequent use of ICRP 68 models to calculate the dose, did not always result in 

the intended highest dose to an organ. Similarly, the appropriate solubility types between the 

two methodologies were not always paired consistently, resulting in discrepancies that were not 

claimant favorable. 

The dose reconstructor is directed to use surrogate radionuclides for radionuclides absent from 

the IMBA code. In lieu of this, a more prudent approach to the absence of radionuclides is to use 

the dose coefficients provided in ICRP 2001 and employ a linear interpolation for the 

radionuclides that are not explicitly given. ORAUT-OTIB-0001 (Brackett 2003) should be 

updated to reflect these and other changes made to the high-five approach. The closure of this 

issue is pending the update of the high-five method using bioassay data and appropriate models. 

It is recommended that issues associated with the high-five approach remain open pending the 

release of the revised ORAUT-OTIB-0001 (Brackett 2003) and subsequent reflection in 

Revision 04-E (Scalsky 2006) of the TBD. 

The issues associated with correction factors and uncertainties have not been satisfactorily 

resolved. It has not been demonstrated that the application of a DAF of 1.119 or 1.039 for both 

TLDs and film for 1952–1986, and an uncertainty of 30% without full consideration of 

laboratory, radiological, and environmental factors, is claimant favorable. The dosimeter 

calibration is based on an incident angle of zero degrees, which underestimates the actual field 

dose where incident angle is greater than zero. The correction factor applied to recorded 

dosimeter results is too low for photon energies from 30 to 250 keV, which is the default photon 

energy used. 

Specific actions recommended by SC&A to achieve closure include the following: 

• Provide detailed period/location film-specific DAFs for the period 1952–1970. 

• Provide period/location TLD-specific DAFs for the period 1971–1986. 

• Assess the impact of new DAFs on coworker data. 

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•	 Account for differences in incident angles between calibration and field use. 

•	 Account for photon energies between 30 and 250 keV (the default photon energy used in 

calibration). 

•	 Clarify the basis for applying a generic 30% uncertainty factor, and/or provide clear 

instructions for applying other appropriate uncertainty factors. 

Revision 03 does not provide additional information to satisfactorily resolve issues associated 

with correction factors and uncertainties. The recent draft of Revision 04E (Scalsky 2006), 

which is not yet officially issued, contains a few changes to external dose reconstruction, but 

these additions do not satisfactorily address issues associated with correction factors and 

uncertainties; therefore, they are still applicable. None of the SRS site-specific workbooks and 

guides that SC&A has been able to locate provides further qualification for using the DAFs 

as recommended in the TBD (Scalsky 2006). The SRS documents that SC&A has been able to 

find do not deal specifically with uncertainties or adjustment factors. In addition, SC&A needs 

more bibliographic information to locate the “guide of 3/29/04” (as listed in the NIOSH 

response) and evaluate its applicability to this issue. It is recommended that for these reasons 

this issue remains open. 

SC&A concurs that ORAUT-OTIB-0017 (Merwin 2005) provides suitable guidance for the 

assignment of shallow dose. This issue is closed. 

The TBD prescribes two very different protocols for neutron dose reconstruction that correspond 

to pre- and post-1971 time periods. Prior to 1971, the uncertainty factors associated with the 

neutron-to-photon ratio are neither technically defensible nor likely to be claimant favorable. 

The TBD provides no compelling evidence to suggest that the TLND dosimeter offered 

significant improvements over NTA film. In brief, this suggests 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 and must be viewed with 

caution. 

SC&A’s evaluation of Revision 03 (Scalsky 2005) did not reveal any changes concerning 

neutron dose reconstruction methodologies. SC&A believes that the use of the geometric mean 

and geometric standard deviation that describe the post-1971 neutron-to-photon ratio is neither 

technically defensible nor likely to be claimant favorable for a large fraction of potential 

claimants. Proposed Revision 04-E of the TBD (Scalsky 2006) provides some changes and 

clarifications to the applications of neutron-to-photon values for dose reconstruction. For likely 

noncompensible cases, the site profile recommends applying the 95th percentile neutron-tophoton dose ratio to the recorded dose. SC&A recommends that the 95th percentile neutron-tophoton values be used in all SRS dose reconstruction cases, not in just the likely 

noncompensable cases. There is currently no Work Group action pending, but it is 

recommended that this issue remain open pending the release of proposed Revision 04-E 

(Scalsky 2006). 

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5.2.3 Objective 3: Adequacy of Data 

The adequacy of the F- and H-Area Tank Farm characterization in the TBD is questionable for 

use as dose reconstruction guidance. Data evaluation appears to be incomplete with regard to 

exposure conditions, radionuclides of concern, and uncertainty. This is particularly true for early 

periods of operation, where primary records involving key operations and incidents are lacking. 

The tank farm database, not currently evaluated by the TBD, can serve to determine what 

assumptions would be suitable in giving claimants who worked in the tank farms the benefit of 

the doubt in the face of considerable uncertainties. The lack of evaluation of primary data 

sources has left the TBD without a realistic way to estimate uncertainties. The potential for 

internal and external exposure to unmonitored workers in areas not designated as radiological 

control areas needs to be investigated. Revision 03 of the site profile made no changes in the 

discussion on the tank farms. Proposed Revision 04-E (Scalsky 2006) made only minimal 

changes, such as the inclusion of Cs-137 and Ru-106 in Table A-14. Default intakes for tank 

farms workers for different periods of time, including actinides, were added. This revision does 

not resolve all issues associated with the tank farms. It is recommended that this issue remain 

open pending completion of Work Group action items and release of the proposed Revision 04-E 

(Scalsky 2006). 

An evaluation of the comprehensiveness of the early monitoring program should be completed 

for early workers to determine whether existing site profile methodologies bound their dose. 

This is especially important in the case of workers who were not monitored but were exposed to 

a radiological hazard. Without a single organization determining neutron dosimeter, bioassay 

requirements, and when special interpretations of film badges are required, there may have been 

inconsistencies in actual practices. The adequacy of the early monitoring program (i.e., who they 

monitored) will not be resolved by an inventory of records provided in claimant files. 

Furthermore, additional validation of the HPAREH database as the exclusive source for external 

coworker dose determination, given the incompleteness of early data, is necessary to demonstrate 

that these data are adequate for this use. It is recommended that this issue remain open pending 

NIOSH’s completion of Work Group action items. 

Revisions 03 and 04-E lack an explanation of why the additional records are not being 

considered for dose reconstruction. SC&A provided NIOSH with an inventory of supplemental 

records that may be beneficial in dose reconstruction. Based on a review of OCAS-PER-019 

(OCAS 2007), NIOSH/ORAUT has received some additional neutron data from SRS. The exact 

nature of this data has not been discussed with SC&A to date. To SC&A’s knowledge, no effort 

has been made by NIOSH/ORAUT to evaluate the completeness of the HPAREH file used in the 

development of the external coworker. The integrity of the HPAREH file for use in coworker 

modeling is questionable, since most of the workers terminating prior to 1979 are not included in 

the HPAREH. A basis for its appropriateness should be developed and discussed in either the 

TBD or the ORAUT-OTIB-0032 (Merwin 2006). Furthermore, consideration should be given to 

multiple dosimetry results, which may provide valuable information on nonuniform exposures 

and may be of assistance in determining organ dose for some workers. For the above reasons, it 

is recommended that this overall finding remain open pending review of the NIOSH Work 

Group action items. 

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5.2.4 Objective 4: Consistency among Site Profiles 

The key assumptions in Revision 03 for medical, environmental, internal, and external dose 

remained the same as those from Revision 02. Section 3.0 of Revision 03 did include an annual 

dose from the ingestion of foodstuffs. Doses are calculated for the ingestion of 60 kilograms 

(wet mass) of food per year for Sr-90, Cs-137, Pu-238, Pu-239, and tritium. NIOSH/ORAUT 

has stated that ingestion is not usually considered at major DOE sites; however, it was 

considered for SRS and for worst-case internal dose assumptions at Hanford. Ingestion dose 

should not be applied selectively at one facility and ignored at another. 

5.2.5 Objective 5: Regulatory Compliance 

The TBD has effectively complied with the hierarchy of data required under 42 CFR Part 82 and 

its implementation guides with one notable exception. SC&A notes that SRS used ICRP 30 to 

determine the relative intakes used in the high-five hypothetical intake. This appears to conflict 

with 42 CFR 82.18(b), which states “NIOSH will calculate the dose to the organ or issue using 

the appropriate current metabolic models published by the ICRP.” 

5.3 USABILITY OF THE SITE PROFILE FOR ITS INTENDED PURPOSES 

SC&A identified a number of issues in the SRS and other site profiles reviewed to date that, in 

some cases, represent potential generic policy issues that transcend any individual site profile. 

These issues may involve the interpretation of existing standards (e.g., oro-nasal breathing), how 

certain critical worker populations (e.g., construction workers and early workers) should be 

profiled for historic radiation exposure, and how exposure itself should be analyzed (e.g., the 

treatment of incidents and statistical treatment of dose distributions). SC&A previously defined 

these issues in its evaluation of Revision 02 of the SRS site profile. NIOSH has issued TIBs to 

address generic issues such as shallow dose assignment, dose to construction workers, exposure 

from highly insoluble plutonium, and exposure from STCs, all of which apply to dose 

reconstruction at SRS. Other common issues still remain, such as the dose from impurities in 

RU, dose to decontamination and decommissioning workers, and quality assurance of records 

provided for claimants by SRS. 

5.3.1 Ambiguous Dose Reconstruction Direction 

The ambiguity of the TBD is currently being addressed under a separate review task. The 

additions and updates to TIBs and procedures have clarified some of the ambiguity in the site 

profile. The use of site workbooks by dose reconstructors ensures that methods are consistent. 

The mingling of data causes confusion in the identification of dose reconstruction 

methodologies. The later revisions of the TBD have made some improvement in this area; 

therefore, it is recommended that this issue be considered closed. 

5.3.2 Inconsistencies and Editorial Errors in the Site Profiles 

NIOSH uses Equations 3-2 and 3-3 on pages 52–53 of the TBD to derive the atmospheric 

dispersion factors (i.e., X/Q values expressed in units of seconds per cubic meter) for groundlevel and elevated releases, respectively. These equations appear to be in error because they 

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result in large X/Q values. For example, using Equation 3-2, the ground-level X/Q at 1,000 

meters downwind from the release point is derived as follows: 

Y = 1.0146X - 1.8809 

where Y = Atmospheric dispersion factor (s/m3) and X = Distance from the source (meters). 

Hence, at 1,000 meters, the X/Q value is as follows: 

Y = 1.0146(1000) - 1.8809 = 1013 

Since X/Q values are typically a small fraction of 1 (e.g., on the order of 0.001), it appears that 

the equation contains a typographical error. Perhaps the equation should be inverted, giving a 

value of 1/1013 or about 0.001. This is also the case for Equation 3-3. 

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6.0 REFERENCES 

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

Occupational Illness Compensation Program Act of 2000, Code of Federal Regulations, May 2, 

2002. 

AEC (U.S. Atomic Energy Commission), 1955, “Intercomparison of Film Badge 

Interpretations,” Isotopics, Volume 2, Number 5, pp. 8–23. 

Allen, D., 2007. The Effect of Additional Neutron Dose Data from the Savannah River Site, 

OCAS-PER-019, Revision 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

May 18, 2007. 

Alvarez, J. et al. 2003. MED/AEC//DOE External Dosimetry Technology Technical Basis 

Document, ORAU, Revision 0, Oak Ridge Associated Universities, Oak Ridge, Tennessee, 

August 21, 2003. 

Baughman, D. F., Hang, P. and Townsend, C. S., Waste Management Tree Data Bank – Status 

Report, WSRC-TR-93-327, August 30, 1993. 

Bebbington, W.P., 1990. History of du Pont at the Savannah River Plant, E. I. du Pont de 

Nemours and Company, Wilmington. 

Beyeler, W.E., W.A. Hareland, F.A. Durán, T.J. Brown, E. Kalinina, D.P. Gallegos, and P.A. 

Davis, 1999. Residual Radioactive Contamination from Decommissioning, Parameter Analysis. 

Draft Report for Comment, NUREG/CR-5512, Vol. 3. U.S. Nuclear Regulatory Commission, 

Washington, DC. 

Bond, R.G., Straub, C.P., and Prober, R. 1972. Air Pollution, Vol. 1 of Handbook of 

Environmental Control. Cleveland, OH: CRC press. 

Boswell, J.M., 2000. "Reactor Production Diversity,” Proceedings of the Symposium May 17, 

2000, 50 Years of Excellence in Science and Engineering at the Savannah River Site, WSRCMS-2000-00061, Westinghouse Savannah River Company, Aiken, South Carolina. 

Brackett, E. M., 2003. Technical Information Bulletin: Maximum Internal Dose Estimates for 

Savannah River (SRS) Claims, ORAUT-OTIB-0001, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee. July 15, 2003. 

Brackenbush, L.W., G.W.R. Endres, J. M. Selby, and E.J. Vallario, 1980. Personnel Neutron 

Dosimetry at Department of Energy Facilities. PNL-3213. Pacific Northwest Laboratory, 

Richland, WA 99352. 

Brackenbush, L. W., Soldat, K. L., Haggard, D. L., Faust, L. G., and Tomeraasen, P. L., 1987. 

Neutron Dose and Energy Spectra Measurements at Savannah River Plant, PNL-6301, Pacific 

Northwest Laboratories, Richland, WA 99352. 

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

October 2 , 2007 

Revision No. 

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Addendum to SCA-TR-TASK1-0003 

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119 of 146 

Brodsky, A., and R.L. Kathren, 1963. Accuracy and Sensitivity of Film Measurements of 

Gamma Radiation –Part I: Comparison of Multiple-Film and Single-Quarterly-Film 

Measurements of Gamma Dose at Several Environmental Conditions, Health Phys 9(4):453-461. 

Brodsky, A., and R.L. Kathren, 1963. Accuracy and Sensitivity of Film Measurements of 

Gamma Radiation –Part II: Limits of Sensitivity and Precision, Health Phys 9(5):463-471. 

Brodsky, A., A.A. Spritzer, F.E. Feagin, F.J. Bradley, G. Karches and H.I. Mandelberg, 1963. 

Accuracy and Sensitivity of Film Measurements of Gamma Radiation –Part IV: Intrinsic and 

Extrinsic Errors, Health Phys 11(10):1071-1082. 

Buschborn, R.L. and E.S. Gilbert. 1993. Summary of Recorded External Radiation Doses for 

Hanford Workers 1944–1989. PNL-8909. Pacific Northwest Laboratory, Richland, 

Washington. 

Carlton, W.H., Bauer, L.R., Evans, A.G., Geary, L.A., Murphy, C.E., Jr., Pinder, J.E., Strom, 

R.N., 1992a. Cesium in the Savannah River Site Environment, WSRC-RP-92-250, 

Westinghouse Savannah River Company, Aiken, South Carolina, March, 1992. 

Carlton, W.H., Evans, A.G., Geary, L.A, Murphy, Jr., C.E., Strom, R.N., 1992b. Assessment of 

Strontium in the Savannah River Site Environment, WSRC-RP-92-984, Westinghouse Savannah 

River Company, Aiken, South Carolina. 

Carlton, W.H., Evans, A.G., Murphy, Jr., C.E., Tuck, D.M., 1993a. Assessment of Radiocarbon 

in the Savannah River Site Environment, WSRC-RP-93-215, Westinghouse Savannah River 

Company, Aiken, South Carolina, March, 1993. 

Carlton, W.H., Denham, M., and Evans, A.G., 1993b. Assessment of Technitium in the 

Savannah River Site Environment, WSRC-RP-93-217, Westinghouse Savannah River Company, 

Aiken, South Carolina. 

Carlton, W.H. and Murphy, Jr., W.H., 1995. Assessment of Noble Gases in the Savannah River 

Site Environment, WSRC-TR-95-219, Westinghouse Savannah River Company, Aiken, South 

Carolina. 

Carlton, W.H. and Murphy, Jr., W.H., 1996. Assessment of Activation Products in the 

Savannah River Site Environment, WSRC-TR-95-0422, Westinghouse Savannah River 

Company, Aiken, South Carolina, July 1996. 

Carlton, W.H., 1999. Dose estimates for PAR pond dam repair, SRT-EXT-99-176, 

Westinghouse Savannah River Company – Office Memorandum, Aiken, South Carolina. 

CDC, 2001. Savannah River Site Dose Reconstruction Project Phase II: Source Term 

Calculation and Ingestion Pathway Data Retrieval, Evaluation of Material Released from the 

Savannah River Site, RAC Report No. 1-CDC-SRS-1999, Centers for Disease Control and 

Prevention, Atlanta, Georgia, April 30, 2001. 

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

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Revision No. 

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Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

120 of 146 

Cheng, Y.S., Dahl, A.R., and Jow, H.N., 1997. Dissolution of Metal Tritides in a Simulated 

Lung Fluid, Health Physics, Vol. 73, Number 4, Williams and Wilkins, Baltimore, Maryland. 

Chew, M. L., Devanney, S. A., Faust, L. G., Morris, R. L., and Potter, E. W., 2006. Parameters 

to Consider when Processing Claims for Construction Trade Workers, ORAUT-OTIB-0052, 

Revision 00, Oak Ridge Associated Universities, Oak Ridge, Tennessee, August 31, 2006. 

Crase, K.W., and LaBone, T. R., 2000. Dose Contribution from Plutonium and Other Impurities 

and Uranium Waste Streams from Savannah River Site Uranium Recovery Facilities, ESH-HPT2000-00040, Westinghouse Savannah River Company, Aiken, South Carolina. 

Cummins, C.L., Martin, D.K., and Todd, J.L., 1991. Savannah River Site Environmental Report 

for 1990 (U). Westinghouse Savannah River Company, WSRC-IM-91-28. Volume II. 

Dennis, R., 1976. Handbook of Aerosols. TID 26608. U.S. ERDA, Washington, DC. U.S. 

ERDA. 

DOE (U.S. Department of Energy) 1985. The Report of the Joint Task Force on Uranium 

Recycle Materials Processing, Oak Ridge Operations, Report No. DOE/OR-859, September 

1985. 

DOE (U.S. Department of Energy) 1994. Department of Energy Radiological Control Manual, 

DOE/EH-0256T, Revision 1, U.S. Department of Energy, Washington, DC. 

DOE (U.S. Department of Energy), 1995. Epidemiological Studies: Guides to Records Series of 

the Department of Energy, “Savannah River Site.” DOE Openness Human Radiation 

Experiments website, www.tis.eh.doe.gov/ohre/new/findingaids/epidemiologic. 

DOE (U.S. Department of Energy) 1996. Plutonium: The First 50 Years, United States 

Plutonium Production, Acquisition, and Utilization from 1944–1994, DOE/DP-0137, 

Department of Energy, Washington, DC. 

DOE (U.S. Department of Energy) 2000. Exposure Assessment Project at the Paducah Gaseous 

Diffusion Plant, Office of Environmental Safety and Health, Washington, DC. December 2000. 

DOE (U.S. Department of Energy) 2001. A preliminary Review of the Flow and Characteristics 

of Recycled Uranium throughout the DOE Complex, F-001-001, March 2001. 

DOE (U.S. Department of Energy) 2004. Radiological Control Programs for Special Tritium 

Compounds, DOE-HDBK-1184-2004, U.S. Department of Energy, Washington, DC. 

DPSOP 1953. Radiation and Contamination Control, Revision 0, DPSOP-40, Savannah River 

Site, Aiken, South Carolina, March, 1953. 

DPSOP 1956. Radiation and Contamination Control, Revision 2, DPSOP-40, Savannah River 

Site, Aiken, South Carolina, January, 1956. 

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DPSP 1956a. Monthly Progress Report, July 1956, Works Technical Department, DPSP-56-1-7, 

Savannah River Site, Aiken, South Carolina, August 16, 1956, p.506. 

DPSP 1956b. Monthly Progress Report, August 1956, Works Technical Department, DPSP-56­

1-8, Savannah River Site, Aiken, South Carolina, September 18, 1956, p. 511. 

DPSP 1956c. Monthly Progress Report, September 1956, Works Technical Department, DPSP56-1-9, Savannah River Site, Aiken, South Carolina, October 18, 1956, p. 513. 

DPSP 1956d. Monthly Progress Report, October 1956, Works Technical Department, DPSP-56­

1-10, Savannah River Site, Aiken, South Carolina, November 15, 1956, p. 517. 

DPSP 1956e. Monthly Progress Report, November 1956, Works Technical Department, DPSP56-1-11, Savannah River Site, Aiken, South Carolina, December 18, 1956, p. 417. 

DPSP 1956f. Monthly Progress Report, December 1956, Works Technical Department, DPSP56-1-12, Savannah River Site, Aiken, South Carolina, January 18, 1957. 

DPSP 1957a. Monthly Progress Report, January 1957, Works Technical Department, DPSP-57­

1-1, Savannah River Site, Aiken, South Carolina, February 18, 1957, p. 410. 

DPSP 1957b. Monthly Progress Report, February 1957, Works Technical Department, DPSP57-1-2, Savannah River Site, Aiken, South Carolina, March 20, 1957, p. 411. 

DPSP 1957c. Monthly Progress Report, April 1957, Works Technical Department, DPSP-57-1­

4, Savannah River Site, Aiken, South Carolina, May 17, 1957, p. 416. 

DPSP 1957d. Monthly Progress Report, May 1957, Works Technical Department, DPSP-57-1­

5, Savannah River Site, Aiken, South Carolina, June 20, 1957, p. 418. 

DPSP 1957e. Monthly Progress Report, June 1957, Works Technical Department, DPSP-57-1-6, 

Savannah River Site, Aiken, South Carolina, July 18, 1957, p. 422. 

DPSP 1957f. Monthly Progress Report, July 1957, Works Technical Department, DPSP-57-1-7, 

Savannah River Site, Aiken, South Carolina, August 19, 1957, p. 419. 

DPSP 1957g. Monthly Progress Report, August 1957, Works Technical Department, DPSP-57­

1-8, Savannah River Site, Aiken, South Carolina, September 19, 1957, p. 421. 

DPSP 1957h. Monthly Progress Report, September 1957, Works Technical Department, DPSP57-1-9, Savannah River Site, Aiken, South Carolina, October 18, 1957, p. 416. 

DPSP 1957i. Monthly Progress Report, October 1957, Works Technical Department, DPSP-57­

1-10, Savannah River Site, Aiken, South Carolina, November 20, 1957, p. 420. 

DPSP 1957j. Monthly Progress Report, November 1957, Works Technical Department, DPSP57-1-11, Savannah River Site, Aiken, South Carolina, December 18, 1957, p. 421. 

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DPSP 1957k. Monthly Progress Report, December 1957, Works Technical Department, DPSP57-1-12, Savannah River Site, Aiken, South Carolina, January 20, 1958, p. 418. 

DPSP 1958a. Monthly Progress Report, June 1958, Works Technical Department, DPSP-58-1-6, 

Savannah River Site, Aiken, South Carolina, July 18, 1958, p.420. 

DPSP 1958b. Monthly Progress Report, July 1958, Works Technical Department, DPSP-58-1-7, 

Savannah River Site, Aiken, South Carolina, August 18, 1958, p. 416. 

DPSP 1958c. Monthly Progress Report, August 1958, Works Technical Department, DPSP-58­

1-8, Savannah River Site, Aiken, South Carolina, September 19, 1958, p.416. 

DPSP 1958d. Monthly Progress Report, September 1958, Works Technical Department, DPSP58-1-9, Savannah River Site, Aiken, South Carolina, October 20, 1958, p. 418. 

DPSP 1958e. Monthly Progress Report, October 1958, Works Technical Department, DPSP-58­

1-10, Savannah River Site, Aiken, South Carolina, November 20, 1958, p. 420. 

DPSP 1958f. Monthly Progress Report, December 1958, Works Technical Department, DPSP58-1-12, Savannah River Site, Aiken, South Carolina, January 23, 1959, p. 4 

DPSOP 1959. Radiation and Contamination Control, Revision 3, DPSOP-40, Savannah River 

Site, Aiken, South Carolina, November, 1959. 

DPSP 1959a. Monthly Progress Report, January 1959, Works Technical Department, DPSP-59­

1-1, Savannah River Site, Aiken, South Carolina, February 19, 1959, p. 417. 

DPSP 1959b. Monthly Progress Report, February 1959, Works Technical Department, DPSP59-1-2, Savannah River Site, Aiken, South Carolina, March 18, 1959, p. 423. 

DPSP 1959c. Monthly Progress Report, March 1959, Works Technical Department, DPSP-59­

1-3, Savannah River Site, Aiken, South Carolina, April 16, 1959, p. 423. 

DPSP 1959d. Monthly Progress Report, April 1959, Works Technical Department, DPSP-59-1­

4, Savannah River Site, Aiken, South Carolina, May 18, 1959, p. 424. 

DPSP 1959e. Monthly Progress Report, May 1959, Works Technical Department, DPSP-59-1-5, 

Savannah River Site, Aiken, South Carolina, June 17, 1959, p. 421. 

DPSP 1959f. Monthly Progress Report, June 1959, Works Technical Department, DPSP-59-1-6, 

Savannah River Site, Aiken, South Carolina, July 21, 1959, p.418. 

DPSP 1959g. Monthly Progress Report, July 1959, Works Technical Department, DPSP-59-1-7, 

Savannah River Site, Aiken, South Carolina, August 19, 1959, p.422. 

DPSP 1959h. Monthly Progress Report, August 1959, Works Technical Department, DPSP-59­

1-8, Savannah River Site, Aiken, South Carolina. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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DPSP 1959i. Monthly Progress Report, September 1959, Works Technical Department, DPSP59-1-9, Savannah River Site, Aiken, South Carolina. 

DPSP 1959j. Monthly Progress Report, October 1959, Works Technical Department, DPSP-59­

1-10, Savannah River Site, Aiken, South Carolina, November 18, 1959, p. 422. 

DPSP 1959k. Monthly Progress Report, November 1959, Works Technical Department, DPSP59-1-11, Savannah River Site, Aiken, South Carolina, December 17, 1959, p. 419. 

DPSP 1959l. Monthly Progress Report, December 1959, Works Technical Department, DPSP59-1-12, Savannah River Site, Aiken, South Carolina, January 20, 1960, p. 423. 

DPSOP 1960. Radiation and Contamination Control, Revision 4, DPSOP-40, Savannah River 

Site, Aiken, South Carolina, June, 1960. 

DPSP 1960a. Monthly Progress Report, January 1960, Works Technical Department, DPSP-60­

1-1, Savannah River Site, Aiken, South Carolina, February 19, 1960, p. 421. 

DPSP 1960b. Monthly Progress Report, March 1960, Works Technical Department, DPSP-60­

1-3, Savannah River Site, Aiken, South Carolina, April 20, 1960, p. 426. 

DPSP 1960c. Monthly Progress Report, April 1960, Works Technical Department, DPSP-60-1­

4, Savannah River Site, Aiken, South Carolina, May 19, 1960, p. 429. 

DPSP 1960d. Monthly Progress Report, May 1960, Works Technical Department, DPSP-60-1­

5, Savannah River Site, Aiken, South Carolina, June 20, 1960, p. 429. 

DPSP 1960e. Monthly Progress Report, June 1960, Works Technical Department, DPSP-60-1-6, 

Savannah River Site, Aiken, South Carolina, July 20, 1960, p. 426. 

DPSP 1960f. Monthly Progress Report, July 1960, Works Technical Department, DPSP-60-1-7, 

Savannah River Site, Aiken, South Carolina, August 17, 1960, p. 423. 

DPSP 1960g. Monthly Progress Report, August 1960, Works Technical Department, DPSP-60­

1-8, Savannah River Site, Aiken, South Carolina, September 20, 1960, p. 429. 

DPSP 1960h. Monthly Progress Report, September 1960, Works Technical Department, DPSP60-1-9, Savannah River Site, Aiken, South Carolina, October 19, 1960, p. 418. 

DPSP 1960i. Monthly Progress Report, October 1960, Works Technical Department, DPSP-60­

1-10, Savannah River Site, Aiken, South Carolina, November 12, 1960, p. 415. 

DPSP 1960j. Monthly Progress Report, November 1960, Works Technical Department, DPSP60-1-11, Savannah River Site, Aiken, South Carolina, December 21, 1960, p. 407. 

DPSP 1960k. Progress Report, December 1960, Works Technical Department, DPSP-60-1-12, 

Savannah River Site, Aiken, South Carolina, January 19, 1961. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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124 of 146 

DPSP 1961. Progress Report, January 1961, Works Technical Department, DPSP-61-1-1, 

Savannah River Site, Aiken, South Carolina, February 16, 1961. 

DPSP 1962a. Progress Report, February 1962, Works Technical Department, DPSP-62-1-2, 

Savannah River Site, Aiken, South Carolina, March 23, 1962. 

DPSP 1962b. Progress Report, May 1962, Works Technical Department, DPSP-62-1-5, 

Savannah River Site, Aiken, South Carolina, June 21, 1962. 

DPSP 1963. Progress Report, February 1963, Works Technical Department, DPSP-63-1-2, 

Savannah River Site, Aiken, South Carolina, March 22, 1963. 

DPSP 1967. Progress Report, March 1967, Works Technical Department, DPSP-67-1-3, 

Savannah River Site, Aiken, South Carolina, April 20, 1967. 

DPSOL 1971. Health Protection Procedures, Section 302, “Bioassay Control,” DPSOL-193­

302, Savannah River Site, Aiken, South Carolina, 1971. 

DPSOL 1987. Health Protection Procedures, Section 213, “Confirmed Assimilation of 

Radionuclides,” DPSOL-193-213, June 1987. 

DPSOL 1988. Health Protection Procedures, Section 212, “Special Bioassay Measurements,” 

DPSOL-193-212, August 1988. 

DPSOL 1989a. Health Protection Procedures, Section 211, “Bioassay Control,” DPSOL-193­

211, March 1989. 

DPSOL 1989b. Health Protection Procedures, Section 207, “Use of Additional Personnel 

Monitoring Devices,” DPSOL-193-211, May 1989. 

DPSOP 1971. Operating Procedures for Radiation and Contamination Control, Revision 37, 

DPSOP-40, Savannah River Site, Aiken, South Carolina, November 1971. 

DPSOP 1974. Operating Procedures for Radiation and Contamination Control, Revision 45, 

DPSOP-40, Savannah River Site, Aiken, South Carolina, January 1974. 

DPSOP 1976. Operating Procedures for Radiation and Contamination Control, Revision 54, 

DPSOP-40, Savannah River Site, Aiken, South Carolina, April, 1976. 

DPSOP 1981. Investigating Radiation and Contamination Incidents, SHB 2, DPSOP-40, 

Savannah River Site, Aiken, South Carolina, March 1981. 

DuPont 1957. Monthly Progress Report, December 1956. DPSP-56-1-12, E.I. du Pont de 

Nemours and Company, Savannah River Plant, Aiken, South Carolina. 

DuPont 1958. Monthly Progress Report, December 1957. DPSP-57-1-12, E.I. du Pont de 

Nemours and Company, Savannah River Plant, Aiken, South Carolina. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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Revision No. 

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Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

125 of 146 

DuPont 1959. Monthly Progress Report, December 1958. DPSP-58-1-12-DV, E.I. du Pont de 

Nemours and Company, Savannah River Plant, Aiken, South Carolina. 

DuPont 1960. Memorandum to H.L. Butler from W.M. Stewart, Subject: Enriched Uranium 

Recycle. October 10, 1960, DPSP-60-2122. 

DuPont 1961. Monthly Progress Report, December 1960. DPSP-60-1-12, E.I. du Pont de 

Nemours and Company, Savannah River Plant, Aiken, South Carolina. 

DuPont 1962. Explosives Department, Atomic Energy Division, Savannah River Plant, Monthly 

Progress Report, Works Technical Department, DPSP series, Savannah River Site Reading 

Room, Aiken, South Carolina. 

DuPont 1965. Progress Report: Works Technical Department, DPSP 65-01-07, Aiken, South 

Carolina, August 1965. 

DuPont 1978. Health Physics Activities at the Savannah River Plant, DPSPU 78-11-17, du Pont 

de Nemours and Company, Savannah River Plant, Aiken, South Carolina. 

DuPont 1990. Dosimetry Special Hazards Investigation Index, FE5300-2000-2004-044-P001, 

E.I. du Pont de Nemours and Company, Savannah River Plant, Aiken, South Carolina. 

Elliott, S.D., 1997. 221/211-H Characterization Plan for Low-Level Radioactive Waste, NMPSHC-94-0971, Revision 8, September 19, 1997. 

ERDA (Energy Research and Development Administration) 1976. Letter from H. Doran 

Fletcher, Director Uranium Enrichment Operations Division, to National Lead Company of Ohio 

and Goodyear Atomic Corporation, Subject: The Handling of Low-Enriched Nitrate Solutions 

from Savannah River. March 8, 1976. 

First, M.W. and Drinker, P. 1952. “Concentrations of Particulates Found in Air.” Arch. Ind. 

Hyg. Occup. Med. 5:387. 

Gilbert, T.L., et al. 1983. “Pathways Analysis and Radiation Dose Estimates for Radioactive 

Residues at Formerly Utilized MED/AEC Sites. ORO-832 (Rev). 

Gorson, R.O., N. Suntharalingam, and J.W. Thomas, 1963. Results of a Film-Badge Reliability 

Study, Presented at the 49th Annual Meeting of the Radiological Society of North America, 

Chicago, Illinois. November 17–22, 1963. 

Gray, P., 2006. High Flux at SRP, electronic message to D. E. Bihl, May 1, 2006. 

Healy, J.W., 1971. Surface Contamination: Decision Levels, LA-4558-MS. Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, September 1971. 

Heederik, D., J. de Cock, and E. Endlich. 1994. Dust exposure indices and lung function 

changes in longshoremen and dock workers. Amer. J. Ind. Med. 26(4):497-509. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

126 of 146 

Hine, G.L. and G.L. Brownell. 1956. Radiation Dosimetry. Academic Press, Inc., New York, 

New York. 

Hoy, J. E. 1972. Personnel Albedo neutron Dosimeter with Thermoluminescent 7Li and 6Li. 

DP-1277. E.I. du Pont de Nemours and Company, Inc., Savannah River Plant, Aiken, South 

Carolina. 

Hoy, J. E. 1980. Thermoluminescent Dosimeters for Personnel Neutron Monitoring. DPST-70­

533. E.I. du Pont de Nemours and Company, Inc., Savannah River Plant, Aiken, South Carolina. 

ICRP 1960. Report of Committee II on Permissible Dose for Internal Radiation, ICRP 

Publication 2, Pergamon Press, Oxford, England. 

IRCP 1977. Recommendations of the International Commission on Radiation Protection, ICRP 

Publication 26, Pergamon Press, Oxford, England. 

IRCP 1979. Limits for the Intake of Radionuclides by Workers, ICRP Publication 30, Pergamon 

Press, Oxford, England. 

ICRP 1990. Recommendations of the International Commission on Radiological Protection. 

ICRP Publication 60, Annals of the ICRP, Vol. 21. Pergamon Press, Oxford, England. 

ICRP 1993. Age-dependent Doses to Members of the Public from Intake of Radionuclides: 

Part 2, Ingestion Dose Coefficients. ICRP Publication 67. Pergamon Press, Oxford, England. 

1993. 

ICRP 1994. Human Respiratory Tract Model for Radiological Protection, ICRP Publication 66, 

Pergamon Press, Oxford, England. 

ICRP 1995. Dose Coefficients for Intakes of Radionuclides by Workers, ICRP Publication 68, 

Pergamon Press, Oxford, England. 

ICRP 1996. Age-Dependent Doses to Members of the Public From Intakes of Radionuclides: 

Part 4 Inhalation Dose Coefficients, ICRP Publication 71, Pergamon Press, Oxford, England. 

ICRP 1996. Conversion Coefficients for Use in Radiological Protection Against External 

Radiation. ICRP Publication 74, Pergamon Press, Oxford, England. 

ICRP 1999. Individual Monitoring for Internal Exposure of Workers, ICRP Publication 78, 

Pergamon Press, Oxford, England. 

ICRP 2001. The ICRP Database of Dose Coefficients: Workers and Members of the Public, 

CD-ROM, Version 2.01, Pergamon Press, Oxford, England. 

Jannik, G.T., 1997. Assessment of SRS Radiological Liquid and Airborne Contaminants and 

Pathways, WSRC-TR-97-0152, Westinghouse Savannah River Company, Aiken, South 

Carolina, April 1997. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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Revision No. 

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Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

127 of 146 

Kantelo, M.V., Bauer, L.R., Marter, W.L., Murphy, Jr., C.E., Zeigler, C.C., 1993. Radiodine in 

the Savannah River Site Environment, WSRC-RP-90-424-2, Westinghouse Savannah River 

Company, Aiken, South Carolina, January 15, 1993. 

Kathren, R.L., Shockley, R. E., Thomas, E. M. and Taulbee, T. D., 2003. Dose Reconstruction 

from Occupationally Related Diagnostic X-ray Procedures, ORAUT-OTIB-0006, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee. 

Kathren, R.L. and Shockley, V.E., 2005. Technical Information Bulletin: Dose Reconstruction 

from Occupationally Related Diagnostic X-ray Procedures, Revision 03 PC-1, ORAUT-OTIB0006, Oak Ridge Associated Universities, Oak Ridge, Tennessee. December 21, 2005. 

Kennedy, W.E. and D.L. Strenge. 1992. Residual Radioactive Contamination from 

Decommissioning: Volume 1, Technical Basis for Translating Contamination levels to Annual 

Total Effective Dose Equivalent, NUREG/CR-5512. Prepared by pacific Northwest Laboratory 

for the U.S. Nuclear Regulatory Commission, Washington, DC, December 1992. 

LaBone, T.R., 1996. Facility Descriptions (U), ESH-HPT-960197, Westinghouse Savannah 

River Company, Savannah River Site, Aiken, South Carolina. 

LaBone, T.R. 2007. Calculation of Dose from Special Tritium Compounds, ORAUT-OTIB0066, Revision 00, Oak Ridge Associated Universities Team, Oak Ridge, Tennessee. April 26, 

2007. 

Linch, KD., 2002. “Respirable concrete dust-silicosis hazard in the construction industry,” Appl 

Occup Environ Hyg Mar; 17(3):209-21, 2002. 

Makhijani, A.B., Alvarez, R., and Blackwelder, B., 1986. Deadly Crop in the Tank Farm, 

Environmental Policy Institute, Washington, DC, 1986. 

McCarty, L.E., 2000. Historical Generation and Flow of Recycled Uranium at the Savannah 

River Site, ESH-PEQ-2000-00059, Westinghouse Savannah River Company, Aiken, South 

Carolina. 

McConville, G.T. and Wood, C.M., 1995. Calculation of Tritium Dose from Insoluble 

Particulates, Fusion Technology: 28:905-909, October 1995. 

Merwin, S.E., 2005. Technical Information Bulletin: Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Revision 00. Oak Ridge Associated 

Universities Team, Cincinnati, Ohio, January 19, 2005. 

Merwin, S., 2006. External Coworker Dosimetry Data for the Savannah River Site, ORAUTOTIB-0032, Revision 00 PC-1, Oak Ridge Associated Universities, November, 7, 2006. 

Milham, R.C. and Boni, A.L., 1976. Detection and Measurement of Tritium Forms Released 

from A Nuclear Production Complex, Proceedings of 24th Conference on Remote System 

Technology, 1976. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

128 of 146 

Miller, D.A. 1956. “Letter to R.C. Blair, Atomic Energy Commission. Subject: Tritium Loss232-F. DPSP-56-966-DV. SRS Phase II Database PGVI994942885. September 17. 

Minnick W. V., and Wellmaker, K.A., 1995. Waste Management Facilities Fault-Tree Data 

Bank 1995, WSRC-TR-0173, Safety Engineering Department, Westinghouse Savannah River 

Corporation, August 16, 1995. 

Morgan, K. Z., 1961. “Dosimetry Requirements for Protection from Ionizing Radiation,” 

Selected Topics in Radiation Dosimetry, Proceedings of the Symposium on Selected Topics in 

Radiation Dosimetry, Sponsored by and Held in Vienna June 7–11, 1960. International Atomic 

Energy Agency, Vienna, Austria, pp. 3–23. 

Morgan, R., 2006. E-mail from Russ Morgan of SRS to K.A. Robertson-DeMers, Re: 

EEOICPA Questions, September 5, 2006. 

Moxley, J., 2002. Letter from J. Moxley to D.P. Potocik, Radiological Control Program for 

Special Tritium Compounds (U), ESH-HPT-2002-00096, Westinghouse Savannah River 

Company, Aiken, South Carolina, June 27, 2002. 

NIOSH (National Institutes of Occupational Safety and Health) 2002a. External Dose 

Reconstruction Implementation Guideline. OCAS-IG-001, Revision 1; August 2002. 

National Council on Radiation Protection and Measurement (NCRP), 1971. Protection Against 

Neutron Radiation. NCRP Report 38. National Council on Radiation Protection and 

Measurement, Bethesda, Maryland. 

Neton, J.W., 2003a. Misinterpreted Dosimetry Records Resulting in an Underestimate of Missed 

Dose in SRS Dose Reconstruction, Revision 0, OCAS-PER-001, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, September 8, 2003. 

Neton, J.W., 2003b. Error in Surrogate Organ Assignment Resulting in an Underestimate of Xray Dose in SRS Dose Reconstructions, Revision 0, OCAS-PER-002, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, December 15, 2003. 

Neton, J.W., 2003c. Neutron Exposures at the Savannah River Site, OCAS-TIB-007, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, September 17, 2003. 

Neton, J.W., 2004. Interpretation of External Dosimetry Records at the Savannah River Site 

(SRS), OCAS-TIB-006, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

Feburary, 20, 2004. 

Nichols, G.N., Vachon, L.J., Tetzlaff, R.N., Waters, D.E., 1954. Special Hazards Incident 

Investigation Class I Number 12, Savannah River Site, Aiken, South Carolina. 

NRC (U.S. Nuclear Regulatory Commission) 1974. Assumptions Used for Evaluating the 

Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water 

Reactors, NRC Regulatory Guide 1.4, Revision 2, June 1974. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

129 of 146 

OCAS 2007. The Effect of Additional Neutron Data from the Savannah River Site, OCAS-PER019, May 18, 2007. 

ORAUT (Oak Ridge Associated Universities Team) 2005. Dose Reconstruction Considerations 

for Recycled Uranium Contaminants, ORAUT-OTIB-0053, draft, Oak Ridge, Tennessee. 

Oztunali, O.I., Re, G.C., Moskowitz, P.M., Picazo, E.D., and Pitt, C.J., 1981. Data Base for 

Radioactive Waste Management, Impacts Analyses Methodology Report. NUREG/CR-1759. 

Washington, DC: U.S. NRC. 

Parker, H.M, 1945. Comparison of Badge Film Readings at the Metallurgical Laboratories, 

Clinton Laboratories and the Hanford Engineering Works, 7-3090, Hanford Atomic Products 

Operation, Richland, Washington. (SRS used Clinton dosimetry services and then implemented 

systems similar to other laboratories) 

Plato, P., 1978. Testing and Evaluating Personal Dosimetry Services in 1976, Health Phys 

34(3):219-223. 

Rando, R.J., Shi, R., Hughes, J.M., Weill, H., McDonald, A.D., and McDonald, J.C., 2001. 

Cohort Mortality Study of North American Industrial Sand Workers. III. Estimation of Past and 

Present Exposures to Respirable Crystalline Silica, Annals of Occupational Hygiene, Volume 45, 

pp. 209-216. 

Reed, M.B., Swanson, M.T., Gaither, S., Joseph, J.W., and Henry, W.R., 2002. Savannah River 

Site at Fifty. U.S Government Printing Office, Washington, DC. 

Richardson D.R., S. Wing, and E. Daniels, 2006. Evaluation of external radiation dosimetry 

records at the Savannah River Site, 1951–1989, Journal of Exposure Science and Environmental 

Epidemiology (2006) 

Robertson-DeMers, K.A., 2003. Description of the External Radiation Monitoring Program at 

the Savannah River Site, Spectrum Technical Services, Kennewick, Washington. Report 

submitted to NIOSH-HERB, March 2003. 

Robertson-DeMers, K.A., 2007. Interview Notes from Conversation with Russ Morgan, 

February 28, 2007. 

Savannah River Site (SRS). 1993. External Dosimetry Technical Basis Manual. WSRC-IM-92­

101, dated January 15, 1993. Westinghouse Savannah River Company, Aiken, South Carolina. 

29808. 

SC&A 2004. SC&A Standard Operating Procedure for Performing Site Profile Reviews, 

Revision 0, March 18, 2004. 

SC&A 2005a. Draft Review of NIOSH Site Profile for Savannah River Site, SCA-TR-TASK1­

0003, S. Cohen and Associates, McLean, Virginia. March 21, 2005. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

Effective Date: 

October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

130 of 146 

SC&A 2005b. The Review of NIOSH/ORAUT Procedures and Methods Used for Dose 

Reconstruction, Report, SCA-TR-Task3-0001, S. Cohen and Associates, McLean, Virginia. 

SC&A 2005c. Review of NIOSH Site Profile for Mallinckrodt Chemical Company, St. Louis 

Downtown Site St. Louis, Missouri, SCA-TR-TASK1-0002, S. Cohen and Associates, McLean, 

Virginia. 

Scalsky, E.D., 2004. Technical Basis Document for the Savannah River Site To Be Used for 

EEOICPA Dose Reconstructions, ORAUT-TKBS-0003, Revision 2, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. 

Scalsky, E.D., 2005. Technical Basis Document for the Savannah River Site to be used for 

EEOICPA Dose Reconstructions, ORAUT-TKBS-0003, Revision 03, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. April 05, 2005. 

Scalsky, E.D., 2006. Technical Basis Document for the Savannah River Site to be used for 

EEOICPA Dose Reconstructions, ORAUT-TKBS-0003, Revision 04-E, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. November 6, 2006. 

Schwendiman, L.C., 1977. Supporting Information for the Estimation of Plutonium Oxide Leak 

Rates Through Very Small Apertures, BNWL-2198. Richland, Washington: Pacific Northwest 

Laboratory. 

Sehmel, G.A., 1977. Radioactive Particle Resuspension Research Experiments on the Hanford 

Reservation, BNWL-2081, UC-11, February 1977. 

Sehmel, G.A., 1980. Particle Resuspension: A Review, Environmental International, Volume 4, 

pp. 107–127. 

Siebert, S.R., 2004. Tritium Calculated and Missed Dose Estimates ,Revision 00, ORAUTOTIB-0011, Oak Ridge Associated Universities Team, Oak Ridge, Tennessee, June 29, 2004. 

Sinclair, P.C., 1976. “Vertical Transport of Desert Particulates by Dust Levels and Clear 

Thermals,” Proceedings of the Atmospheric-Surface Exchange of Particulates and Gaseous 

Pollutants– 1974 Symposium, pp 497–527, USERDA, CONF-740921, NTIS, Springfield, 

Virginia. 1976. 

Stern, A.C., 1976. “Air Pollution,” In: Air Pollutants, Their Transformation and Transport. 

Volume 1. New York: Academic Press. 

Stewart, K., 1964. “The Resuspension of Particulate Material from Surfaces,” in Surface 

Contamination, Proceedings of a Symposium Held at Gatlinburg, Tennessee, June 1964. 

Pergamon Press, Oxford, 1967. 

Sutter, S.L., 1982. Accident Generated Particulate Materials and Their Characteristics – A 

Review of Background Information, NUREG/CR-2651. Battelle Pacific Northwest Labs., 

Richland, Washington. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

131 of 146 

Sweet, C. W. and Murphy, Jr. C.E., 1982. Tritium Deposition in Pine Trees and Soil From 

Atmospheric Releases of Molecular Tritium, E.I. DuPont de Nemours and Company, Aiken, 

South Carolina. 

Taylor, G. A., K.W. Crase, T.R. LaBone, and W.H. Wilkie, W.H., 1995. A History of Personnel 

Radiation Dosimetry at the Savannah River Site, Westinghouse Savannah River Company, 

WSRC-RP-95-234, Aiken, South Carolina. 

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis and P. Andreo, 2002. Study of a Selection 

of 10 Historical Types of Dosemeter: Variation of the Response of HP(10) with Photon Energy 

and Geometry of Exposure. Radiat Prot Dos, 102(2): 101-113. 

Thomas, H.E. and D.A. Phoenix, 1976. Summary Appraisals of the Nations Ground-Water, 

Geological Survey Professional Paper, 1976 (a series of separate documents addressing 

individual regions of the U.S., each with a different series number). 

Till, J.E. and H.R. Meyer, 1983. Radiological Assessment Division of System Integration, Office 

of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC. 

UPA (United Power Association) 1974. Final Elk River Reactor Program Report, COO-651-93. 

Elk River, Minnesota. 

Vallario, E. J, D.E. Hankins and C.M. Unruh. 1969. AEC Workshop on Personnel Neutron 

Dosimetry. BNWL-1340. Battelle, Pacific Northwest Laboratory, Richland, Washington. 

Vallario, E.J., D.E. Hankins and C.M. Unruh. 1971. Second AEC Workshop on Personnel 

Neutron Dosimetry. BNWL-1616. Battelle, Pacific Northwest Laboratory, Richland, 

Washington. 

Watson, Jr., J.E., J.L. Wood, W.G. Tankersley, and C.M. West, 1994. Estimation of Radiation 

Doses for Workers without Monitoring data for Retrospective Epidemiologic Studies. Health 

Physics. 67(4): 402-405. 

Wellmaker, K. A., 1995. Determination of the Frequencies of Tritium Releases and Fire Related 

Incidents for Tritium Facilities SAR (U), WSRC-TR-95-2062. 

Wilson, R.H., J.J. Fix, W.V. Baumgartner, and L.L. Nichols. 1990. Description and Evaluation 

of Hanford Personnel Dosimeter Program From 1944 Through 1989. PNL-7447. Pacific 

Northwest Laboratory, Richland, Washington. 

WSRC 1995. Risk Assessment Methodology Fault Tree Databank – User Guide (U), WSRCIM-95-50. 

WSRC 1996. Assessment of Selected Fission Products in the Savannah River Site Environment, 

WSRC-TR-96-0220, Westinghouse Savannah River Company, Aiken, South Carolina. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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WSRC 1997. Assessment of Neptunium, Americium, and Curium in the Savannah River Site 

Environment, WSRC-TR-97-00266, Westinghouse Savannah River Company, Aiken, South 

Carolina. 

WSRC 1999. Independent Assessment of the Savannah River Site Internal Dosimetry Program, 

ESH-HPT-99-0022, Westinghouse Savannah River Company, February 9, 1999. 

WSRC, 2000, 20 Years of Excellence in Science and Engineering at the Savannah River Site, 

Proceedings of the Symposium May 17, 2000, WSRC-MS-2000-00061, Westinghouse Savannah 

River Company, Aiken, South Carolina. 

WSRC 2001. Internal Dosimetry Technical Basis Manual, WSRC-IM-90-139, Westinghouse 

Savannah River Company, Aiken, South Carolina. 

Yu et al. 1993. Data Collection Handbook to Support Modeling the Impacts of Radioactive 

Material in Soil. ANL/EAIS-8. Environmental Assessment and Information Division, Argonne 

National Laboratory. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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ATTACHMENT 1: NIOSH TECHNICAL DOCUMENTS CONSIDERED 

DURING THE REVIEW PROCESS 

Technical Basis Documents: 

• ORAUT-TKBS-0003, Savannah Rive Site, Revision 03, April 5, 2005 (Scalsky 2005). 

Technical Support Documents: 

•	 OCAS-PER-001, Misinterpreted Dosimetry Records Resulting in an Underestimate of 

Missed Dose in SRS Dose Reconstruction, Revision 0, Office of Compensation Analysis 

and Support, Cincinnati, Ohio, September 8, 2003. (Neton 2003a) 

•	 OCAS-PER-002, Error in Surrogate Organ Assignment Resulting in an Underestimate of 

X-ray Dose in SRS Dose Reconstructions, Revision 0, Office of Compensation Analysis 

and Support, Cincinnati, Ohio, December 15, 2003. (Neton 2003b) 

•	 OCAS-PER-0019, The Effect of Additional Neutron Dose Data from the Savannah River 

Site, Revision 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, May 

18, 2007. (Allen 2007) 

•	 OCAS-TIB-006, Interpretation of External Dosimetry Records at the Savannah River 

Site (SRS), Revision 1, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

February 20, 2004. (Neton 2004) 

•	 OCAS-TIB-007, Neutron Exposures at the Savannah River Site, Revision 0, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, September 17, 2003. (Neton 

2003c) 

•	 ORAUT-OTIB-0001, Technical Information Bulletin: Maximum Internal Dose 

Estimates for Savannah River Site (SRS) Claims, Revision 0, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, July 15, 2003. (Brackett 2003) 

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

Revision 0 PC-1, Oak Ridge Associated Universities, Oak Ridge, Tennessee, November 

7, 2006. (Merwin 2006) 

•	 ORAUT-OTIB-0006, Revision 02 (2003), Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-ray Procedures, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee. December 29, 2003. (Kathren et al. 

2003) 

•	 ORAUT-OTIB-0006, Revision 03 (2005), Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-ray Procedures, Oak Ridge 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

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Associated Universities, Oak Ridge, Tennessee. December 21, 2005. (Kathren and 

Shockley, 2005) 

•	 ORAUT-OTIB-0011, Revision 00, (2004b) Technical Information Bulletin – Tritium 

Calculated and Missed Dose Estimates, Oak Ridge Associated Universities, Oak Ridge, 

Tennessee, June 29, 2004. (Siebert 20204) 

•	 ORAUT-OTIB-0017, Revision 00 (2005), Technical Information Bulletin – 

Interpretation of Dosimetry Data for Assignment of Shallow Dose, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, January 19, 2005. (Merwin 2005) 

•	 ORAUT-OTIB-0066 (2007), Revision 00, Calculation of Dose from Intakes of Special 

Tritium Compounds, Oak Ridge Associated Universities, Oak Ridge, Tennessee, April 

26, 2007. (LaBone 2007) 

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ATTACHMENT 2: WSMS DATABANK STATISTICS 

SRS staff obtained a copy of the WSMS incident database. Because this database was lengthy 

and contained classified information, a review team was assembled to review the database at 

SRS from February 28 through March 1, 2007. The three objectives were to determine the 

contents of the database, compare entries in this database to those from the tank farm Fault Tree 

Databank, and determine its usefulness in dose reconstruction. The WSMS database contains 

464,092 incidents, including many that do not relate to radiation exposure. SRS has expressed 

concern about having to review the entire database for classified material and has requested that 

the Work Group identify the particular incidents of interest. To accommodate uncleared 

members of SC&A, NIOSH, ORAUT, and the Work Group, incident counts by year, area code, 

and facility code have been provided. Counts for key equipment codes and operations codes 

associated with sample incidents retrieved from the database have also been provided. However, 

without the user manual, SC&A was unable to define many of the codes listed below. 

Table A2-1. WSMS Incident Statistics by Year 

Year # of Incidents Year # of Incidents 

1953 1 1980 18708 

1954 107 1981 18600 

1955 301 1982 18689 

1956 415 1983 16879 

1957 426 1984 18571 

1958 390 1985 19745 

1959 653 1986 25366 

1960 597 1987 23944 

1961 624 1988 20402 

1962 752 1989 24538 

1963 804 1990 27456 

1964 1308 1991 36531 

1965 1368 1992 31525 

1966 1918 1993 31124 

1967 2102 1994 32240 

1968 1633 1995 28057 

1969 1203 1996 5292 

1970 1454 1997 38 

1971 1971 1998 47 

1972 2783 1999 80 

1973 2578 2000 28 

1974 3018 2001 10 

1975 3571 2002 5 

1976 5163 2003 19 

1977 8831 2004 35 

1978 9856 2005 5 

1979 12324 2006 1 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

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Table A2-2. WSMS Incident Statistics by Area Code 

Area ID Count Area Code 

19 6 

V (ON SITE OTHER THAN 

IDENTIFIED AREAS) 

15 22 R 

20 22 W 

17 33 T 

10 52 L 

13 64 P 

21 76 Z 

3 90 C 

7 94 G 

2 105 B 

12 168 N 

9 185 K 

4 347 D 

18 384 V 

5 613 E 

14 800 Q 

16 1867 S 

1 22131 A 

11 27816 M 

8 157644 H 

6 251710 F 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

 

 

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Table A2-3. WSMS Incident Statistics by Facility Code 

Facility ID Facility Code Count 

1 SA 1 

2 A 3991 

3 B 381 

4 C 856 

5 D 207 

6 E 725 

7 F 109 

8 H 197 

9 I 667 

10 K 47 

11 LA 45 

12 LB 1335 

13 LC 19 

14 LD 330 

15 LE 18855 

16 LF 5 

17 LG 292 

18 LH 1 

19 LI 36 

20 LJ 1 

21 LK 125 

22 LL 1213 

23 LM 241 

24 LS 73 

25 LT 4 

26 LU 24 

27 LV 21913 

28 LW 46 

29 LY 3 

30 M 3 

31 MC 49 

32 MF 27447 

33 O 1 

34 Q 303 

35 S 233 

36 S0 25 

37 SA 170555 

38 SB 120668 

39 SC 36473 

40 SD 15970 

41 SE 5503 

Facility ID Facility Code Count 

42 SG 2 

43 SH 93 

44 SI 7443 

45 SJ 495 

46 SK 119 

47 SL 19454 

48 SM 3365 

49 SN 88 

50 SO 183 

51 SP 1048 

52 SQ 28 

53 SS 19 

54 ST 13 

55 SW 784 

56 SX 486 

57 SZ 355 

58 T 151 

59 W 43 

60 WA 285 

61 WB 27 

62 WC 69 

63 WD 3 

64 WE 39 

65 WG 117 

66 WH 2256 

67 WJ 302 

68 WK 6 

69 WL 13 

70 WM 180 

71 WO 17 

72 WQ 2257 

73 WR 631 

74 WS 1589 

75 WT 21964 

76 WU 335 

77 WV 18 

78 WW 781 

79 WX 13 

80 WY 2 

81 WZ 79 

82 X 1 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

 

 

 

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Table A2-4. WSMS Incident Statistics by Key Equipment Codes Related to 

Incidents Involving Radiation 

Equipment ID Equipment Code Count 

8 313-M (FABRICATION) 13352 

62 ACCEPTANCE CRITERIA 4 

63 ACCOUNTABILITY 667 

70 ACTIVITY 2386 

74 AIR EMMISSIONS 35 

76 AIR REVERSAL 740 

80 AIRBORNE ACTIVITY (772-F) 597 

83 ALARM/HORN 33979 

90 AMERICIUM 29 

91 AMERICIUM, CURIUM, CALIFORNIUM 62 

94 ANALYSIS 165 

95 ANALYTICAL 1893 

96 ANALYTICAL CELL 867 

97 ANALYTICAL, 320-M 130 

108 ASSAY/MONITOR ENRICHED URANIUM 7 

163 BETA-GAMMA INCINERATOR 118 

182 BIOASSAY AND/OR CHEST COUNT 263 

183 BIOASSAY AND/OR CHEST COUNT (772-F) 33 

190 BLOWER, FANS 11380 

192 BODY EXPOSURE>2RAD/HR OR>1R/HR 186 

205 BREATHING AIR 3237 

212 BUILDING 232-H 67 

213 BUILDING 234-H 43 

214 BUILDING 235-H 3 

215 BUILDING 238-H 2 

216 BUILDING 249-H 2 

219 BULDING 233-H 29 

222 BURIAL GROUND 1600 

225 CABINET 147 

226 CABINET, ENTRY/EXIT OF MATERIAL 4386 

230 CALIBRATION 3084 

231 CALIBRATION SOURCE 495 

232 CALIFIRNIUM FACILITY 53 

297 CLEANING 541 

302 CLOTHING CONTAMINATION (772-F) 738 

337 CONDUCT OF OPERATIONS 344 

345 CONTAMINATION 2123 

346 CONTAMINATION PERSONAL CLOTHING/EFFECTS 395 

347 CONTAMINATION, AIRBORNE 19160 

348 CONTAMINATION, CLOTHING 4977 

349 CONTAMINATION, FACILITY OR EQUIPMENT 29421 

350 CONTAMINATION, NASAL 706 

351 CONTAMINATION, SKIN 2475 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

 

 

 

 

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Table A2-4. WSMS Incident Statistics by Key Equipment Codes Related to 

Incidents Involving Radiation 

Equipment ID Equipment Code Count 

355 CONTROL ROOM/CONTROL BOARD/PANEL BOARD 2190 

368 CORE MACHINING 99 

374 CORE, URANIUM 35 

380 CRITICALITY POTENTIAL 2099 

386 CURIE (0 TO 1E2) 75 

387 CURIE (1E2 TO 1E3) 871 

388 CURIE (1E3 TO 1E4) 79 

389 CURIE (1E4 TO 1E5) 77 

390 CURIE (1E5 TO 1E6) 75 

391 CURIUM 242 8 

392 CURIUM 244 10 

398 DAMPER 925 

407 DECONTAMINATION 1 

418 DEMOLITION/DECOMMISION 8 

419 DENITRATOR 54 

466 DRUMS, CANS 979 

469 DUCT 1220 

476 DUST BAG 23 

477 DUST COLLECTION 244 

480 EFFLUENT TREATMENT FACILITY 377 

496 ENRICHED URANIUM METAL RECEIVING, 321-M 68 

497 ENRICHED URANIUM STORAGE 162 

498 ENVIRONMENTAL PROTECTION AGENCY 1269 

499 ENVIRONMENTAL RELEASE 2476 

500 ENVIRONMETAL PROGRAMS, MISC. 300-AREA 1250 

504 EQUIPMENT SEALS 416 

506 ERRORS 1797 

507 ERRORS, SUPERVISOR 882 

509 ESTIMATED TOTAL EXPOSURE (772-F) 158 

510 EU CONCENTRATE TANK 36 

511 EU LOADOUT FACILITY 231 

513 EVACUATION 482 

517 EVAPORATOR (GENERAL) 176 

535 FACILITY CONTAMINATION (772-F) 1762 

548 FILTERS (GENERAL), SCREENS, STRAINER 9652 

560 FIRE 2116 

561 FIRE DETECTION AND SUPPRESSION 18059 

562 FIRE EXTINGUISHER 13 

563 FIRE WATCH 114 

564 FIRE WATER SYSTEM 158 

567 FIXED CONTAMINATION 184 

582 FLUOROSCOPE 119 

583 FLUOROSCOPE INSPECTION, 321-M 119 

603 FULL BODY AND RESPIRATORY PROTECTION (772-F) 1291 

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

 

 

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Table A2-4. WSMS Incident Statistics by Key Equipment Codes Related to 

Incidents Involving Radiation 

Equipment ID Equipment Code Count 

604 FUMES 605 

614 GANG VALVE 6713 

614 GANG VALVE 6713 

631 GLOVES 11839 

632 GLOVES (772-F) 355 

638 GROUND WATER 281 

639 GROUNDWATER CLEANUP 684 

654 HAZARDOUS WASTE 865 

656 HEAT EXCHANGER/COOLER 819 

662 HEPA FILTER 1163 

675 HOOD OR RADIOBENCH (772-F) 552 

676 HOOD/GLOVE BOX 541 

681 HOSE 1693 

691 HUT 1403 

702 IDENTIFICATION, URANIUM SLUGS 34 

709 IMPROPER STORAGE 799 

710 IMPURITIES 782 

714 INADEQUATE ADMINISTRATIVE CONTROL 392 

715 INADEQUATE COMMUNICATION 1818 

718 INADEQUATE MONITORING 37 

719 INADEQUATE PROTECTIVE CLOTHING 805 

726 INDUSTRIAL HYGIENE 102 

730 INHALATION 6767 

731 INJURY 36 

734 INJURY, MEDICAL TREATMENT CASE 33 

743 INSTALL, REPLACE (AND/OR REMOVE) 7841 

744 INSTRUMENT AIR 3468 

745 INSTRUMENT MALFUNCTION 68255 

752 INTERLOCK 1316 

760 INVENTORY 186 

761 IODINE 155 

762 IODINE REACTOR 1317 

763 ION CHAMBER 8 

764 ISOTOPE SEPARTION (GAS) 37 

772 KANNE MONITOR 483 

775 LABORATORY 631 

786 LEAKS 37354 

795 LIMIT EXCEEDED 1221 

800 LIQUID EFFLUENT TREATMENT FACILITY (LETF) 674 

801 LIQUID LEVEL 2485 

819 LOSS 854 

823 LOW LEVEL WASTE 190 

844 MAINTENANCE ROOM/AREA 69 

850 MASS SPECTROMETER 83 

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

 

 

 

 

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Table A2-4. WSMS Incident Statistics by Key Equipment Codes Related to 

Incidents Involving Radiation 

Equipment ID Equipment Code Count 

853 MATERIAL LOST 49 

868 METER 186 

872 MISC 300 AREA FACILITIES 5350 

874 MISLABELED 974 

877 MIXED WASTE 46 

882 MONITORING 3591 

890 NEPTUNIUM 261 

891 NEPTUNIUM TUBE FABRICATION, 321-M 16 

892 NEUTRON MONITOR 1578 

893 NEUTRON TEST GAGE, 321-M 262 

919 OPERATING ERROR 3255 

920 

OPERATIONAL SAFETY REQUIREMENT/TECH SAFETY 

REQUIREMENT 436 

954 PERSONAL EXPOSURE 1656 

955 PERSONNEL 4513 

958 PERSONNEL EXPOSURE>20MR (772-F) 500 

960 PERSONNEL RADIATION MONITOR 29 

969 PIPE/LINES 6860 

976 PLUGGAGE 42235 

977 PLUTONIUM 1268 

978 PLUTONIUM 238 13 

979 PLUTONIUM STORAGE FACILITY 2284 

997 PRECIPITATOR 6 

992 PRE-EXTRUSION (U-AL LOGS), 321-M 148 

1012 PROCEDURAL DIFFICULTY 3118 

1013 PROCEDURAL VIOLATION 2986 

1014 PROCEDURE, INADEQUATE 868 

1016 PROCESS CONTAMINATION MONITOR (CAM, STORM WATER) 2315 

1020 PROCESS MONITORS 80 

1021 PROCESS RADIAITON MONITOR (VAMPS, ETC.) 627 

1026 PROTECTIVE CLOTHING 2316 

1029 PRSONNEL CONTAMINATION MONITOR 627 

1036 PUNCTURE 321 

1041 QUALITY CONTROL 427 

1052 RADIATION 5 

1053 RADIO BENCH 45 

1054 RADIOACTIVE 1155 

1055 RAIL/TRACK 305 

1060 RDZ, RCA, RBA 137 

1069 RECEIVING PU 238 9348 

1093 RELEASE GUIDE 3 

1105 RESPIRATORY PROTECTION 415 

1116 RISER 16 

1133 RUPTURED 1374 

1134 RUTHENIUM 68 

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

 

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Table A2-4. WSMS Incident Statistics by Key Equipment Codes Related to 

Incidents Involving Radiation 

Equipment ID Equipment Code Count 

1139 SAFETY EQUIPMENT 270 

1150 SAMPLING/SAMPLER 13369 

1180 SEWER 795 

1186 SHIELDING 76 

1201 SKIN CONTAMINATION (772-F) 224 

1209 SLUGS, URANIUM 37 

1213 SMEARABLE CONTAMINATION 542 

1225 SOLID BURNING 9 

1228 SOLVENT BURNING 19 

1243 SPILL 2850 

1251 STACK 1936 

1252 STACK MONITOR 365 

1266 STEP OFF PAD 16 

1280 STRUCTURAL COMPONENT 2173 

1285 SUMP 18872 

1288 SURVEILLANCE 750 

1364 TARGET FABRICATION FACILITY 23 

1368 TECHNICAL DIVISION (SRTC) 1849 

1380 THERMAL PROCESSING PU 238 12 

1385 THORIUM 137 

1400 TRAINING 660 

1417 TRANSPLUTONIUM 138 

1426 TRITIUM OXIDE 2 

1427 TRITIUM, D20 251 

1428 TROUBLE ALARM/TROUBLE LIGHT 656 

1454 UNCONTROLLED REACTION 483 

1463 UPTAKE/ABSORPTION 15 

1464 UPTAKE/INGESTION 13 

1465 UPTAKE/INHALATION 397 

1466 UPTAKE/INJECTION 74 

1467 URANIUM 766 

1476 VALVE FAILURE 26181 

1479 VAULT 1555 

1483 VENTILATION 4578 

1484 VENTILATION (772-F) 2408 

1486 VENTILATION, TANK 802 

1489 VESSEL/TANK/CONTAINER 9386 

1505 WASTE 2459 

1508 WASTE DISPOSAL 2152 

1509 WASTE DISPOSAL (772-F) 680 

1528 WELL WATER 475 

1529 WELLS 630 

1535 X-RAY 74 

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Table A2-5. WSMS Incident Statistics by Key Operations Codes Related to Incidents 

Involving Radiation 

Operation ID Operation Description Count 

4 313-M SLUG FABRICATION OPERATIONS – 300 FACILITY 12209 

5 320-M ROD ASSEMBLY OPERATIONS – 300 FACILITY 1702 

6 321-M TUBULAR ASSEMBLY OPERATIONS – 300 FACILITY 4942 

7 322-M SLUG & TUBULAR ASSEMBLY OPERATIONS – 300 FACILITY 7807 

10 ACID RECOVERY 3171 

11 ACTINIDE TARGET FABRICATION 791 

12 ADJUSTMENT/ION EXCHANGE (NP-237) 2367 

13 ADJUSTMENT/PRECIPITATION/FILTRATION (PU-38) 2898 

15 ANION EXCHANGE COLUMN 8869 

17 BASIN WATER CLEANUP/WATER OPERATIONS 476 

18 BETA-GAMMA INCINERATOR 7 

23 CABINETS 856 

24 CASK OPERATIONS 612 

28 CHEMICAL PREPARATION AND STORAGE 635 

29 CHEMICAL STORAGE 3354 

30 CHEMICAL TRANSFER FACILITY (CTF) 645 

35 COMPRESSED GASES 7998 

37 CONSTRUCTION 1513 

39 CRANE AND HOIST OPERATIONS 18681 

40 CRITICALITY 808 

44 DENITRATOR, A-LINE 4861 

51 ELECTRICAL 13830 

53 ENVIRONMENTAL 4218 

55 EVAPORATOR 3178 

57 EXTENDED SLUDGE PROCESSING (ESP) 183 

58 F.P. REMOVAL FROM EVAPORATOR CONDENSATE 18 

59 FILTRATION 7722 

61 FINISHING/PACKAGING/PU-238 OXIDE 418 

64 FIRES 17567 

67 FUEL AND TARGET OPERATIONS 529 

69 FUEL STORAGE 230 

71 GANG VALVE CORRIDOR 10306 

73 GAS-TRITIUM DATABANK 2 

79 HEALTH PROTECTION 4893 

83 INSPECTIONS/TESTS 867 

85 LABORATORY (TRITIUM FACILITY) 25 

86 LAUNDRY 324 

87 MECHANICAL OPERATIONS 573 

88 MECHANICAL PROCESSES (IN CELL) 75 

89 MISCELLANEOUS 34045 

90 MPPF FINISHING AND HANDLING 27 

91 MPPF PROCESS CONTROL 55 

92 MPPF SEPARATION PROCESSES 92 

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

Effective Date: 

October 2 , 2007 

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Table A2-5. WSMS Incident Statistics by Key Operations Codes Related to Incidents 

Involving Radiation 

Operation ID Operation Description Count 

93 MPPF SOLUTION PREPARATION 51 

95 NEW PRODUCTION REACTOR (NPR) 15 

96 NEW SPECIAL RECOVERY 8351 

104 OVERFLOWS 1048 

110 PERSONNEL SAFETY 1314 

114 PRECIPITATION/FILTRATION (NP-237) 3406 

116 PROCESS CONTROL 1929 

118 PROCESS OPER (TRITIUM) 81 

119 PROCESS SAFETY 978 

120 PRODUCT STORAGE AND ACCOUNTABILITY 737 

123 QUALITY ASSURANCE 13 

132 SAMPLE AISLE 11235 

133 SAMPLING 1276 

135 SECOND PRODUCT CYCLE 5810 

136 SECOND URANIUM CYCLE 3186 

142 SHIPPING/RECEIVING/STORAGE (NP-237) 333 

144 SLAG AND CRUCIBLE 155 

145 SOLIDS HANDLING, A-LINE 1654 

147 SOLVENT WASHING 4378 

148 SPECIAL RECOVERY 8577 

150 STORAGE/BASIN OPERATIONS 659 

153 TANK 5176 

154 TANKS 353 

155 THERMAL PROCESSES (NP-237) 221 

169 UNCONTROLLED REACTIONS 311 

176 VENTILATION 22447 

180 WASTE – HAZARDOUS 831 

181 WASTE – INTERMEDIATE LEVEL 110 

182 WASTE – LOW LEVEL 188 

183 WASTE – MIXED 64 

185 WASTE – TRU 97 

187 WASTE HANDLING 10002 

188 WATER HANDLING FACILITIES 1083 

189 WATER SYSTEMS 15884 

190 WET CHEMISTRY (SCRAP RECOVERY, HBLINE) 3351 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

145 of 146 

ATTACHMENT 3: RECOMMENDED CRITERIA FOR RETRIEVAL OF 

INCIDENTS FROM WSMS 

As a result of the February 28–March 1, 2007, review of the data bank, the Work Group asked 

SC&A to provide the results of several queries run during the database review to obtain some 

general statistics on the database. Attachment 2 of this report summarizes the results. SC&A 

was asked to recommend criteria to the working group that would isolate the incidents of interest 

in the WSMS database. These criteria will be reviewed in a future Work Group meeting and 

modified by NIOSH, ORAUT, and the Work Group as necessary. Incidents are categorized by 

year, area, facility, source of information, equipment, and operations. It is recommended that the 

incidents retrieved first be limited by year and include incidents occurring prior to 1990. By 

1990, incidents should have been filed regularly in the individual dose record. It is 

recommended that no restriction be placed on the area identification (ID) and the facility ID. 

Source ID, equipment ID, and operations ID, which are numerical values, can be used to focus 

the search to incidents involving radioactive material. The numerical values for each ID of 

interest are outlined below. Since the equipment ID and operations ID descriptions for an 

incident often contain text, it is suggested that these fields be searched for keywords. Suggested 

keywords are listed below. While visiting SRS, the review team printed more than 100 incidents 

from the database and cleared for release. SC&A chose the specific source ID, equipment ID, 

operations ID, and keywords based on the occurrence of the values and keywords in the 100 

example incidents. In addition, if the ID did not occur in the incident record and it was clearly of 

interest (e.g., uptake/ingestion), the ID has been included below. These criteria constitute a 

preliminary recommendation for consideration by the Work Group in further meetings. 

The following search terms are recommended to obtain records representative of the types of 

incidents of interest from the WSMS database. 

Incidents occurring in 1990 or before 

And, 

Incidents where the equipment field contains the words uptake, intake, inhalation, ingestion, 

health protection, environment, contamination, uranium, neptunium, bioassay, chest count, 

americium, californium, curie, curium, fume, fire, limit exceeded, neutron, exposure, plutonium, 

thorium, tritium, stack, alarm, or x-ray. 

Or, 

Incidents where the operations field contains the words miscellaneous, health protection, special 

recovery, fires, or ventilation. 

Or, 

Incidents where Source ID equals 2, 12, 46, 47, 55, 61, 62, 63, 64, 69, 74, 76, 101, 112, 114, 121, 

122, 123, 124, 126, 127, 130, 131, 145, 150, 156, 157, 160, 174, 176, 179, 182, 190, 191, 214, 

225, 233, 248, 262, 263, 271, 277, 296, 309, 310, 321, 322, 323, 331, 341, 376, 380, 381, or 386. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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

Effective Date: 

October 2 , 2007 

Revision No. 

Draft 

Document No. 

Addendum to SCA-TR-TASK1-0003 

Page No. 

146 of 146 

Or, 

Incidents where equipment ID equals 8, 62, 63, 70, 74, 76, 80, 83, 90, 91, 94–97, 108, 163, 182, 

183, 190, 192, 205, 212, 213, 214, 215, 216, 219, 222, 225, 226, 230, 231, 232, 297, 302, 337, 

345–351, 355, 368, 374, 380, 386–392, 398, 407, 418, 419, 466, 469, 476, 477, 480, 496–500, 

504, 506, 507, 509, 510, 511, 513, 517, 535, 548, 560–564, 567, 582, 583, 603, 604, 614, 631, 

632, 638, 639, 654, 656, 662, 675, 681, 691, 702, 709, 710, 714, 715, 718, 719, 726, 730, 731, 

734, 743, 744, 745, 752, 760, 761, 762, 763, 764, 772, 775, 786, 795, 800, 801, 819, 823, 844, 

850, 853, 868, 872, 874, 877, 882, 890, 891, 892, 893, 919, 920, 954, 955, 958, 960, 969, 976, 

977, 978, 979, 992, 997, 1012, 1013, 1014, 1016, 1020, 1021, 1026, 1029, 1036, 1041, 1052, 

1053, 1054, 1055, 1060, 1069, 1093, 1105, 1116, 1133, 1134, 1139, 1150, 1180, 1186, 1201, 

1209, 1213, 1225, 1228, 1243, 1251, 1252, 1228, 1243, 1251, 1252, 1266, 1280, 1285, 1288, 

1364, 1368, 1380, 1385, 1400, 1417, 1426, 1427, 1428, 1454, 1463, 1464, 1463, 1464, 1465, 

1466, 1467, 1476, 1479, 1483, 1484, 1486, 1489, 1505, 1508, 1509, 1528, 1529, or 1535. 

Or, 

Incidents where operations ID equals 4, 5, 6, 7, 10, 11, 12, 13, 15, 17, 18, 23, 24, 28, 29, 30, 35, 

37, 39, 40, 44, 51, 53, 55, 57, 58, 59, 61, 64, 67, 69, 71, 73, 79, 83, 85, 86, 87, 88, 89, 90, 91, 92, 

93, 95, 96, 104, 110, 114, 116, 118, 119, 120, 123, 132, 133, 135, 136, 142, 144, 145, 147, 148, 

150, 153, 154, 155, 169, 176, 180, 181, 182, 183, 184, 185, 187, 188, 189, or 190. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and 

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