January 25, 2006 

David Staudt 

Contracting Officer 

Centers for Disease Control & Prevention 

Acquisition and Assistance Field Branch 

P.O. Box 18070 

626 Cochrans Mill Road – B-140 

Pittsburgh, PA 15236-0295 

Subject: Contract No. 200-2004-03805, Task Order 1: Draft Review of the 

NIOSH Site Profile for the Idaho National Laboratory, Idaho – Revision 1 

Dear Mr. Staudt: 

S. Cohen & Associates (SC&A, Inc.) is pleased to submit its revised draft Review of the NIOSH 

Site Profile for the Idaho National Laboratory, Idaho. This report includes Attachment 3, 

“Summaries of Site Expert Reviews,” that was missing from Revision 0 of the document 

submitted to the Advisory Board and NIOSH on September 23, 2005. Revision 1 also includes 

an Attachment 5, “Issue Resolution Matrix for Findings and Key Observations.” Both of these 

attachments have undergone review by DOE. In preparing this deliverable, the authors have also 

made some minor editorial changes that do not affect the content of the original material. 

Sincerely, 

John Mauro, PhD, CHP 

Project Manager 

cc: 

P. Ziemer, PhD, Board Chairperson J. Broehm, NIOSH 

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

L. Wade, PhD, NIOSH K. Behling, SC&A 

L. Elliott, NIOSH D. Chan, PhD, SC&A 

J. Neton, PhD, NIOSH J. Lipzstein, PhD, SC&A 

S. Hinnefeld, NIOSH A. Makhijani, PhD, SC&A 

Z. Homoki-Titus, NIOSH S. Ostrow, PhD, SC&A 

A. Brand, NIOSH J. Fitzgerald, Saliant 

L. Shields, NIOSH K. Robertson-DeMers, CHP, Saliant 

Project File (ANIOS/001/05) 
Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute of Occupational Safety and Health 

Review of the NIOSH Site Profile for the 

Idaho National Laboratory, Idaho
 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0005 

Prepared by 

S. Cohen & Associates 

6858 Old Dominion Drive, Suite 301 

McLean, Virginia 22101 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

January 25, 2006 

Disclaimer 

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

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

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

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

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

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

regarding its conclusions are unwarranted. 
 

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S. COHEN & ASSOCIATES: 

Technical Support for the Advisory Board on 

Document No. 

SCA-TR-TASK1-0005 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program 

Effective Date: 

Draft – January 25, 2006 

Revision No. 1 (Draft) 

REVIEW OF THE NIOSH SITE PROFILE FOR THE 

IDAHO NATIONAL LABORATORY SITE, IDAHO Page 2 of 249 

Task Manager: _________________ Date:____________ 

 Joe Fitzgerald 

Supersedes: 

Rev. 0 (Draft) 

Project Manager: _______________ Date:____________ 

 John Mauro 
 

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TABLE OF CONTENTS 

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

1.0 Executive Summary ...........................................................................................................14 

1.1 Summary of Strengths............................................................................................22 

1.2 Summary of Findings.............................................................................................24 

1.3 Opportunities for Improvement .............................................................................27 

2.0 Scope and Introduction ......................................................................................................30 

3.0 Assessment Criteria and Method .......................................................................................33 

3.1 Objectives ..............................................................................................................33 

3.1.1 Objective 1: Completeness of Data Sources .............................................33 

3.1.2 Objective 2: Technical Accuracy..............................................................33 

3.1.3 Objective 3: Adequacy of Data.................................................................33 

3.1.4 Objective 4: Consistency Among Site Profiles.........................................34 

3.1.5 Objective 5: Regulatory Compliance ........................................................34 

3.2 Assessment Methodology ......................................................................................35 

4.0 Site Profile Strengths .........................................................................................................40 

4.1 Completeness of Data Sources...............................................................................40 

4.2 Technical Accuracy/Claimant-Favorability...........................................................41 

4.3 Adequacy of Data ..................................................................................................42 

4.4 Consistency Among the Site Profiles ....................................................................44 

4.5 Regulatory Compliance .........................................................................................44 

5.0 Vertical Issues ....................................................................................................................45 

5.1 Discussion of Issues...............................................................................................45 

5.1.1 Occupational Environmental Dose Issues .................................................45 

5.1.2 Occupational Internal Dosimetry Issues ....................................................65 

5.1.3 SL-1 Accident Dose Reconstructions ........................................................80 

5.1.4 Occupational External Dose Issues............................................................87 

5.2 Other Observations ..............................................................................................111 

6.0 Overall Adequacy of the INL Site Profile as a Basis for Dose Reconstruction ..............113 

6.1 Satisfying the Five Objectives .............................................................................113 

6.1.1 Objective 1: Completeness of Data Sources ...........................................115 

6.1.2 Objective 2: Technical Accuracy............................................................117 

6.1.3 Objective 3: Adequacy of Data...............................................................119 
 

 

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6.1.4 Objective 4: Consistency Among Site Profiles.......................................120 

6.1.5 Objective 5: Regulatory Compliance ......................................................120 

6.2 Usability of Site Profile for Intended Purpose .....................................................121 

6.3 Unresolved Policy or Generic Technical Issues ..................................................123 

7.0 References........................................................................................................................125 

Attachment 1: Conference Call with NIOSH and SC&A...........................................................131 

Attachment 2: Advance Questions for Site Expert Interviews ...................................................154 

Attachment 3: Summary of Site Expert Interviews ....................................................................166 

Attachment 4: Consistency Between Hanford, SRS, and INL Site Profiles...............................220 

Attachment 5: Issue Resolution Matrix for Findings and Key Observations ..............................240 
 

 

 

 

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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission 

AEDE Annual Effective Dose Equivalent 

α Alpha Particle 

AMWTP Advanced Mixed Waste Treatment Program 

ANC Aerojet Nuclear Corporation 

ANL Argonne National Laboratory 

ANL-W Argonne National Laboratory West 

ANP Aircraft Nuclear Propulsion 

ANPP Aircraft Nuclear Propulsion Program 

ANSI American National Standards Institute 

APS Atmospheric Protection System 

ARA Army Reactor Area (later Auxiliary Reactor Area) 

AREA Army Reactor Experimental Area 

ARMF-1 Advanced Reactivity Measurements Facility No. 1 

ARMF-2 Advanced Reactivity Measurements Facility No. 2 

ARVFS Army Reentry Vehicle Facility Site 

ATR Advanced Test Reactor 

ATRC Advanced Test Reactor Critical Facility 

β Beta Particle 

BORAX Boiling Water Reactor Experiment 

BORAX I Boiling Water Reactor Experiment No. 1 

BORAX II Boiling Water Reactor Experiment No. 2 

BORAX III Boiling Water Reactor Experiment No. 3 

BORAX IV Boiling Water Reactor Experiment No. 4 

BORAX V Boiling Water Reactor Experiment No. 5 

Bq Becquerel (1 disintegration per second) 

Celsius/Centigrade 

CA Contamination Area 

CAM Continuous Air Monitor 

CAMU Corrective Action Management Unit 

CAS Criticality Accident System 

C 
 

 

 

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CDC Centers for Disease Control and Prevention 

CE Critical Experiment 

CEDE Committed Effective Dose Equivalent 

CEL Chemical Engineering Laboratory 

CERCLA Comprehensive Environmental Response Compensation and Liability Act 

CERT Controlled Environmental Radioiodine Tests (later Radionuclides) 

CET Critical Experiment Tank 

CFA Central Facilities Area 

Ci Curie 

cm Centimeter 

cpm Counts per Minute 

CPP Chemical Processing Plant 

D&D Decontamination and Decommissioning 

DAC Derived Air Concentration 

DE Dose Equivalent 

DL Detection Limit 

DNFSB Defense Nuclear Facilities Safety Board 

DOD U.S. Department of Defense 

DOE U.S. Department of Energy 

DOE-HQ DOE-Headquarters 

DOE-ID DOE-Idaho Operations Office 

DOELAP DOE Laboratory Accreditation Program 

DU Depleted Uranium 

EBOR Experimental Beryllium Oxide Reactor 

EBR Experimental Breeder Reactor 

EBR-I Experimental Breeder Reactor No. 1 

EBR-II Experimental Breeder Reactor No. 2 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

EMDR Environmental Monitoring Data Report 

EMR Environmental Monitoring Report 

EPA Environmental Protection Agency 

ERDA Energy Research and Development Administration 
 

 

 

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ESRF Environmental Science and Research Foundation 

ETF Effluent Treatment Facility 

ETR Engineering Test Reactor 

ETRC Engineering Test Reactor Critical 

EXCES Experimental Cloud Exposure Study 

F Fahrenheit 

FARET Fast Reactor Test 

FASB Fuel Assembly and Storage Building 

FAST Fluorinel Dissolution Process and Fuel Storage Facility 

FCF Fuel Cutting Facility 

FDF Fluorinel Dissolution Facility 

FEBT Fuel Element Burn Test 

FECF Fuel Element Cutting Facility 

FFTF Fast Flux Test Facility 

FMF Fuel Manufacturing Facility 

FNCF Facility Neutron Correction Factor 

FP Fission Product 

FPF Fuel Processing Facility 

FPFRT Fission Product Field Release Test 

FPR Fuel Processing Restoration 

FSF Fuel Storage Facility 

γ Gamma 

GCRE Gas-Cooled Reactor Experiment 

GE General Electric Corporation 

GE-ANP General Electric-Advanced Nuclear Propulsion (Program) 

H&S Health & Safety (management organization) 

HCA High Contamination Area 

HEPA High-Efficiency Particulate Air 

HEU Highly Enriched Uranium 

HFEF Hot Fuel Examination Facility 

HLLW High-Level Liquid Waste 

HLW High-Level Waste 
 

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HOTCE Hot Critical Experiment 

HP Health Physicist/Health Physics 

HPIL Health Physics Instrument Laboratory 

HPP Hot Pilot Plant 

HRA High Radiation Area 

HSL Health Services Laboratory 

HTRE Heat Transfer Reactor Experiment 

IBM International Business Machines 

IC Initial Criticality 

ICPP Idaho Chemical Processing Plant (formerly CPP and now INTEC) 

ICRP International Commission on Radiological Protection 

ICRU International Commission on Radiation Units and Measurements 

ID Idaho 

IDO Idaho Operations Office 

IET Initial Engine Test 

IFR Integral Fast Reactor 

IFSF Irradiated Fuel Storage Facility 

ILTSF Intermediate-Level Transuranic Storage Facility 

IMBA Internal Modular Bioassay Analysis 

in Inch 

INC Idaho Nuclear Corporation 

INEC Idaho Nuclear Energy Commission 

INEEL Idaho National Engineering and Environmental Laboratory 

INEL Idaho National Engineering Laboratory 

INELHDE Idaho National Engineering Laboratory Historical Dose Evaluation 

INL Idaho National Laboratory 

INTEC Idaho Nuclear Technology and Engineering Center (formerly ICPP and CPP) 

IRC INEL Research Center (a facility in Idaho Falls) 

IREP Interactive RadioEpidemiological Program 

ISFSI Independent Spent Fuel Storage Installation 

ISU Idaho State University 

IWP Industrial Waste Pond 
 

 

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KERMA Kinetic Energy Released in Matter 

keV Kilo Electron Volt, 1,000 Electron Volts 

kg Kilogram 

kV Kilovolt 

kVp Kilovolt Peak 

kW Kilowatt 

lat Lateral 

lb Pound 

LET Linear Energy Transfer 

LLD Lower Limit of Detection 

LMFBR Liquid Metal Fast Breeder Reactor 

LOCA Loss-of-Coolant Accident 

LOFT Loss of Fluid Test (Facility) 

LPT Low-Power Test (Facility) 

LSC Liquid Scintillation Counter 

M Moderate (solubility rate) 

mA Milliampere 

MDA Minimum Detectable Activity 

MDF Materials Development Facility 

MeV Mega Electron Volt, 1 Million Electron Volts 

mg Milligram 

ML-1 Mobile Low-power Reactor No. 1 

mm Millimeter 

MPBB Maximum Permissible Body Burden 

MPLB Maximum Permissible Lung Burden 

MPOB Maximum Permissible Organ Burden 

mR Milliroentgen 

mrad Millirad 

mrem Millirem 

MRL Minimum Reporting Level 

msec Millisecond 

MTR Materials Test Reactor 
 

 

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MW Megawatt 

NASA National Aeronautics and Space Administration 

NBS National Bureau of Standards 

NCRP National Council on Radiation Protection and Measurement 

NERP National Environmental Research Park 

NIOSH National Institute of Occupational Safety and Health 

NOAA National Oceanic and Atmospheric Administration 

NRAD Neutron Radiography Facility 

NRC Nuclear Regulatory Commission 

NRF Naval Reactor Facility 

NRTS National Reactor Testing Station 

NTA Nuclear Track Emulsion-Type A 

NVLAP National Voluntary Laboratory Accreditation Program 

NWCF New Waste Calcining Facility 

OCAS (NIOSH) Office of Compensation Analysis and Support 

OMRE Organic Moderated Reactor Experiment 

ORAU Oak Ridge Associated Universities 

ORNL Oak Ridge National Laboratory 

PA Posterior-Anterior 

PAS Personal Air Sampler 

PBF Power Burst Facility 

PIC Pocket Ionization Chamber (i.e., "Pencil" Dosimeter) 

PIF Process Improvement Facilities 

POC Probability of Causation 

PREPP Process Experimental Pilot Plant 

psi Pounds Per Square Inch 

R Roentgen 

RAC Risk Assessment Corporation 

rad Radiation Absorbed Dose 

RAF Remote Analytical Facility 

RAL Remote Analytical Laboratory 

RaLa Radioactive Lanthanum 
 

 

 

 

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RAM Radiation/Remote Area Monitor 

RBE Relative Biological Effectiveness 

RBOF Receiving Basin for Offsite Fuel 

RCRA Resource Recovery and Conservation Act 

RCT Radiological Control Technician 

RDT Relative Diffusion Test 

rem Roentgen Equivalent Man 

rep Roentgen-equivalent-physical 

RESL Radiological and Environmental Services Laboratory 

RHA Radiation Hazards Analysis 

RLWTF Radioactive Liquid Waste Treatment Facility 

RM Radioactive Material 

RMF Reactivity Measurement Facility 

RML Radiation Measurements Laboratory 

RPP Radiological Protection Program 

RPSSA Radioactive Parts Service and Storage Area 

RSAC Radiological Safety Analysis Computer 

RSWF Radioactive Scrap and Waste Facility 

RU Recycled Uranium 

RWMC Radioactive Waste Management Complex 

RWMIS Radioactive Waste Management Information Service 

RWP Radiological Work Permit 

S Slow (solubility rate) 

SC&A S. Cohen & Associates 

SDA Subsurface Disposal Area 

SEC Special Exposure Cohort 

SID Source to Image Distance 

SIS Special Isotope Separations 

SL-1 Stationary Low-Power Reactor No. 1 

SM-1 Stationary Medium-Power Reactor 

SMC Specific Manufacturing Capability 

SNAP Systems for Nuclear Auxiliary Power 
 

 

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SNAPTRAN Systems for Nuclear Auxiliary Power Transient 

SPERT Special Power Excursion Reactor Test 

SPF Sodium Processing Facility 

SRS Savannah River Site 

SSD Source to Skin Distance 

SSSTF Staging, Storage, Sizing, and Treatment Facility 

STEP Safety Test Engineering Program 

STM Stack Tritium Monitor 

STP Sewage Treatment Ponds 

STPF Shield Test Pool Facility 

STR Split-Table Reactor 

Sv Sievert 

SWDF Solid Waste Disposal System 

SWEPP Stored Waste Examination Pilot Plant 

SWP Safe Work Permit 

t Thermal 

TAN Test Area North 

TBD Technical Basis Document 

TCE Trichlorethylene 

TEPC Tissue Equivalent Proportional Counter 

TLD Thermoluminescent Dosimeter 

TLND Thermoluminescent Neutron Dosimeter 

TMI Three Mile Island 

TRA Test Reactor Area 

TREAT Transient Reactor Test Facility 

TRIGA Training Research and Isotope General Atomic 

TRU Transuranics 

TRUPACT Transuranic Packaging Transporter 

TSA Transuranic Storage Area 

TSF Technical Support Facility 

UNH Uranium Nitrate Hexahydrate 

U.S.C. United States Code 
 

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USGS U. S. Geological Survey 

WAC Waste Acceptance Criteria 

WBRR Western Beam Research Reactor 

WCA Waste Characterization Area 

WCF Waste Calcining Facility 

WERF Waste Experimental Reduction Facility 

WIPP Waste Isolation Pilot Plant 

WROC Waste Reduction Operations Complex 

WRRTF Water Reactor Research Test Facility 

ZPPR Zero Power Plutonium Reactor (later Zero Power Physics Reactor) 

ZPR III Zero Power Reactor No. 3 

ZPR Zero Power Reactor 
 

 

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1.0 EXECUTIVE SUMMARY 

S. Cohen and Associates (SC&A, Inc.) evaluated the following Technical Basis Documents 

(TBDs) documents related to historical occupational exposures at the Idaho National Laboratory 

(INL1) Site: 

• 	 ORAUT-TKBS-0007-1, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Introduction (Rohrig 2004i) 

• 	 ORAUT-TKBS-0007-2, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Site Description (Rohrig 2005) 

• 	 ORAUT-TKBS-0007-3, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational Medical Dose (Rohrig 2004m) 

• 	 ORAUT-TKBS-0007-4, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational Environmental Dose (Peterson 

2004) 

• 	 ORAUT-TKBS-0007-5, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational Internal Dose (Rich and Wenzel 

2004) 

• 	 ORAUT-TKBS-0007-6, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational External Dosimetry (Rohrig 

2004e) 

These six documents TBDs taken together are often referred to as the INL Site Profile. 

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

(TIBs), the general National Institute for Occupational Safety and Health (NIOSH) dose 

evaluation guidelines, and a supplemental document that relate to the INL Site Profile: 

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

Estimates for Certain DOE Complex Claims (Rollins 2004) 

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

Occupationally Related Diagnostic X-Ray Projections (Kathren 2003) 

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

Ambient Levels of External Radiation (Strom 2003) 

1

The laboratory recently changed names to the Idaho National Laboratory (INL) from its previous name, 

the Idaho National Engineering and Environmental Laboratory (INEEL). Both names will be used in this report to 

refer to the same organization depending on the context. 
 

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• 	 ORAUT-OTIB-0009, Technical Information Bulletin in Support of INEEL Technical 

Basis Document Section 6: Reanalysis of Hankins MTR Bonner Sphere Surveys (Rohrig 

2004h) 

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

(Merwin 2005) 

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

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

• 	 Supplement to Technical Basis Document 4 for the Idaho National Engineering and 

Environmental Laboratory: INEEL Occupational Environmental Dose (Peterson 2004s) 

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

interviews with groups of former and current site personnel (summarized in Attachments 2 and 

3), review of U.S. Department of Energy (DOE) Tiger Team findings, review of Defense 

Nuclear Facilities Safety Board (DNFSB) assessments, some documents originally provided to 

NIOSH by INL staff for the development of these six TBDs, some site documents provided to 

the SC&A team by INL staff during the site interview meetings, and written and teleconference 

communications with personnel from NIOSH’s prime contractor for the dose reconstruction 

project, Oak Ridge Associated Universities (ORAU), and its subcontractors (summarized in 

Attachment 1). The review also examined the popular (i.e., non-scholarly) history book written 

by Susan Stacy for the DOE in recognition of the 50th anniversary of the laboratory: Proving the 

Principle – A History of the Idaho National Engineering and Environmental Laboratory 1949– 

1999 (Stacy 2000). 

The TBDs, supplementing individual claimant exposure data provided by DOE and information 

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

reconstructions under the Energy Employees Occupational Illness Compensation Program Act of 

2000 (EEOICPA). The TBDs contain compilations and analyses of data, such as those related to 

facility operations and processes, radiological source term characterizations, chemical and 

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

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

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

As the support contractor to the Advisory Board on Radiation and Worker Health (Advisory 

Board), SC&A has been charged with independently evaluating the approach taken in NIOSH 

site profiles to gauge their adequacy, completeness, and validity. These evaluations will be used 

by the Advisory Board to advise the Secretary of Health and Human Services on the scientific 

validity, quality, and accuracy of dose reconstruction efforts performed by NIOSH and its 

contractors. 

The INL site occupies 890 square miles (572,000 acres) in southeast Idaho. It consists of nine 

primary facility areas situated on an expanse of otherwise undeveloped, high-desert terrain. 

Buildings and structures are clustered within these primary facility areas, which are typically less 
 

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than a few square miles in size and separated from each other by miles of mostly undeveloped 

land. The site was established in 1949 as the National Reactor Testing Station (NRTS). The 

initial missions at INL were the development of civilian and defense nuclear reactor technologies 

and management of spent nuclear fuel. Fifty-two reactors were built, most of them first-of-akind, experimental devices, including the Navy’s first prototype nuclear propulsion plant. Of 

those 52 reactors, 3 remain in operation today. 

In 1951, INL achieved one of the most significant scientific and engineering accomplishments of 

the 20th century when the Experimental Breeder Reactor No. 1 (EBR-1) became the first facility 

to produce a usable quantity of electricity based on nuclear fission. The EBR-1 is now a 

Registered National Historic Landmark open to the public. During the 1970s, INL’s name was 

changed from the National Reactor Testing Station to the Idaho National Engineering Laboratory 

(INEL) to reflect its broadened mission into areas like biotechnology, energy and materials 

research, and conservation and renewable energy. INL’s name changed again in the spring of 

1997 to the Idaho National Engineering and Environmental Laboratory (INEEL) to reflect a 

major refocus of the laboratory toward engineering applications and environmental solutions. 

On February 1, 2005, the name changed once again to its current name, the Idaho National 

Laboratory (INL). This latest change reflects a renewed mission to support the nation’s 

expanding nuclear energy initiatives (including advanced Generation IV nuclear energy systems, 

nuclear energy/hydrogen co-production technology, and advanced nuclear energy fuel cycle 

technologies) and to support the security needs of the nation’s critical infrastructure. 

INL’s historical role in the Federal nuclear complex included: 

• 	 A nuclear reactor research, development, and testing site; INL operated 52 reactors for 

such applications as electricity generation, naval reactor propulsion, aircraft propulsion, 

and space missions. 

• 	 Nuclear materials production and processing for military and civilian applications. 

• 	 Transuranic waste storage and disposal, primarily from the Rocky Flats plutonium 

foundry in Colorado. INL currently stores the nation’s largest inventory of transuranic 

wastes (about 65,000 m3

 (DOE 2000)). 

• 	 Nuclear reactor fuel research, development, fabrication, and testing. 

• 	 Spent nuclear reactor fuel handling and storage from U.S. and foreign research reactors, 

commercial nuclear power plants, and naval propulsion reactors. 

• 	 Processing and fabrication of depleted uranium armor for the U.S. Army. 

• 	 Environmental restoration, waste management, decontamination, and decommissioning. 

The INL site is composed of nine individual facilities, largely isolated from one another by 

design in the interest of safety. The primary facility areas are (1) the Argonne National 

Laboratory-West, (2) the Central Facilities Area, (3) the Idaho Nuclear Technology and 

Engineering Center, (4) the Naval Reactors Facility, (5) the Waste Reduction Operations 

Complex/Power Burst Facility, (6) the Radioactive Waste Management Complex, (7) the Test 

Area North, (8) the Test Reactor Area, and (9) the Auxiliary Reactor Area. The remainder of the 

site land is referred to as the Sitewide Area, which comprises all INL land outside the boundaries 
 

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of the primary facility areas. In addition, several INL laboratories and administrative offices are 

located in the city of Idaho Falls, some 25 miles east of the INL site boundary. The INL site was 

and remains a complex operation involved in numerous nuclear-related missions, many of them 

experimental, each of which has its own unique exposure hazards. Occupational risks of 

exposure to ionizing radiation are generally defined by INL’s past and current missions. These 

missions are summarized for eight of the nine primary facility areas, omitting the Naval Reactors 

Facility, which operated under separate administration: 

(1) Argonne National Laboratory-West (ANL-W) was established in 1957. The original 

mission was to test nuclear reactors and reactor safety systems, including the 

Experimental Breeder Reactor No.1 (EBR-I) and Boiling Water Reactor Experiments 

(BORAX I-V). The current mission includes stabilization, management, and storage of 

spent nuclear fuel; storage of transuranic waste; and large-scale advanced reactor 

development. Facilities at ANL-W include EBR-II, Fuel Conditioning Facility (FCF), 

Hot Fuel Examination Facility (HFEF), Zero Power Physics Reactor (ZPPR), Transient 

Reactor Test Facility (TREAT), and Sodium Processing Facility. 

(2) Central Facility Area (CFA), established in the 1940s, predating the nuclear laboratory, 

covers a large area, and includes 72 buildings. Its original mission was to lodge U.S. 

Navy gunnery range personnel during World War II, and provide centralized support for 

the INL (1950s–present). The current mission includes treatment and disposal of nonhazardous commercial/industrial waste and support for other INL facilities 

(administrative offices, research laboratories, cafeteria, medical services, construction/ 

support services, workshops, warehouses, landfills, etc.). Facilities at CFA include the 

Hot Laundry Facility, DOE Laboratory Accreditation Procedure Irradiation Facility, 

Health Physics Instrument Laboratory, Handling and Open Storage Area, Remote Service 

Facilities, Administrative Offices and Support Area, Service Shops Area, Radiological 

and Environmental Sciences Laboratory (RESL), Fire Station and Fire Fighting Training 

Facility (FFTF), Light Laboratory, Warehousing and Storage, and INL Sanitary Landfill. 

(3) Idaho Nuclear Technology and Engineering Center (INTEC), formerly Idaho Chemical 

Processing Plant (ICPP or CPP), was established in the early 1950s. In this report, 

INTEC, ICPP, and CPP designations are used interchangeably. Its original mission was 

to reprocess spent nuclear fuels by chemically separating out the reusable uranium (until 

1992) and calcine high-level waste (until 2000). The current mission includes storage of 

low-level, mixed low-level, and high-level radioactive waste and spent nuclear fuel, and 

development of treatment methods for high-level radioactive waste. Facilities at INTEC 

include Fluorinel Dissolution Process and Fuel Storage Facility (FAST), Remote 

Analytical Laboratory, Wet and Dry Fuel Storage Facility, TMI-2 Independent Spent 

Fuel Storage Installation (ISFSI), High-Level Liquid Waste Underground Storage Tank 

Farm, New Waste Calcining Facility, and INTEC Processing Corridors. The old, 

entombed, Waste Calcining Facility is still present. 

(4) Waste Reduction Operations Complex/Power Burst Facility (WROC/PBF) was 

established in the late 1950s. Its original mission was to perform research on small 

power reactors, and to investigate and promote reactor safety. The current mission 
 

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includes storage of spent nuclear fuel, treatment and storage of mixed low-level and lowlevel waste, and research to reduce hazardous and mixed waste. This area housed the 

Special Power Excursion Reactor Tests (SPERT) reactors. Current facilities at 

WROC/PBF include the Power Burst Facility (PBF, formerly SPERT-I), Waste 

Engineering Development Facility/WROC Lead Storage Facility (formerly SPERT-II), 

Waste Experimental Reduction Facility (WERF) (formerly SPERT-III), and Mixed 

Waste Storage Facility (formerly SPERT-IV). 

(5) Radioactive Waste Management Complex (RWMC) was established in late 1952. Its 

original mission was to provide solid, low-level radioactive waste disposal; burial of 

transuranic waste and hazardous substances, such as organic and inorganic chemicals 

(until 1970); storage of transuranic waste on pads above ground; and disposal of other 

waste in 20 pits, 58 trenches, and 21 soil vault rows (1970 to present). The current 

mission calls for interim storage of transuranic waste and shipment of stored transuranic 

waste to WIPP for permanent disposal. Facilities at RWMC include Subsurface Disposal 

Area (SDA), Intermediate Level Transuranic Storage Facility, Transuranic Storage Area 

(TSA), Stored Waste Examination Pilot Plant, TRUPACT Loading Station, and 

Advanced Mixed Waste Treatment Project (AMWTP). 

(6) Test Area North (TAN) was established in 1951, originally for the Aircraft Nuclear 

Propulsion Program (ANP). Later, its mission included investigating core material from 

the damaged Three Mile Island-II reactor, test reactors and nuclear fuel, and 

manufacturing operations. The current mission includes inspection and storage of spent 

nuclear fuel and manufacturing depleted uranium armor for military vehicles (at the 

Specific Manufacturing Facility). Facilities at TAN include ANP, Loss of Fluid Test 

Reactor (LOFT), Three Mile Island Unit 2 (TMI-2) Core Offsite Examination Program, 

Technical Support Facility, Initial Engine Test Facility (IET), Specific Manufacturing 

Capability Program (formerly Containment Test Facility), and Water Reactor Research 

Test Facility. There are other support facilities at TAN, including Hot Shop, Warm Shop, 

Hot Cells, Storage Pool, Storage Pads, Radioactive Liquid Waste Disposal System, 

Radioactive Parts Service and Storage Area, and Radiography Facility. 

(7) Test Reactor Area (TRA) was established in 1982. Its original mission was to study the 

effect of radiation on materials, fuels, and equipment using seven reactors, especially the 

Materials Test Reactor (MTR: 1952–1970), the Engineering Test Reactor (ETR: 1957– 

1982), and the Advanced Test Reactor (ATR: 1967–present). The current mission 

includes wet storage of spent nuclear fuel, operation of the ATR, research supporting the 

U.S. Navy and other customers, and production of isotopes for medicine and industry. 

Other formerly and currently operating facilities at TRA include ATR Critical Facility, 

TRA Hot Cell Facility, TRA Gamma Facility, Radiation Measurements Laboratory, 

Radiochemistry Laboratory, Safety and Tritium Applications Research Facility, Liquid 

Waste Disposal Ponds, and High-Level Liquid Waste Tanks and Transfer Facility. 
 

 

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(8) Auxiliary Reactor Area (ARA), originally called the Army Reactor Experimental Area 

(AREA), was established in 1958 and ceased operations in the 1990s. Its original 

mission was to test stationary, portable, or mobile reactors of low, medium, or high 

power. ARA is located 10 miles east of the Central Facility Area. It began with the 

ARA-1 site, then ARA-II, -III, and -IV at half-mile intervals: ARA-1 housed a hot cell 

facility; ARA-II was renamed the Stationary Low Power Reactor Number 1 (SL-1), 

which was shut down due to an excursion accident; ARA-III was the site for the Army 

Gas Cooled Reactor Experiment (GCRE); and ARA-IV was the site for the Mobile Low 

Power Reactor (ML-1). 

There are other, smaller facilities and experiments located elsewhere on the vast INL site. These 

facilities include the Organic Moderated Reactor Experiment (OMRE), INL Research Center, 

Army Reentry Vehicle Facility Station, and Test Grid III. At Test Grid III, INL housed several 

important experimental facilities, including Fuel Element Burn Tests, Fission Products Field 

Release Tests, Relative Diffusion Tests, and Experimental Cloud Exposure Study. 

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

review to examine in detail all aspects of the site profile, due to the immense complexity and 

long history of the INL facilities, and the many changes that have occurred over the decades. 

SC&A has selected only certain issues for detailed consideration and discussion, because they 

may significantly affect dose reconstruction, and, ultimately, determination of a claimant’s 

petition for compensation. 

Based upon the SC&A Standard Operating Procedure for Performing Site Profile Reviews 

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

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

(1) Completeness of data sources 

(2) Technical accuracy 

(3) Adequacy of data 

(4) Consistency among site profiles 

(5) Regulatory compliance 

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

review, this report discusses the strengths of the TBDs, followed by a description of the major 

“issues” identified during our review. The 35 issues were carefully assessed with respect to the 

five review criteria listed above. Seventeen of the issues are designated as “findings,” because 

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

which have the potential to substantially impact at least some dose reconstructions.2 The 

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

TBDs could improve on. 

2

 Several findings are combined, resulting in eleven distinct findings. 
 

 

 

 

 

 

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Issues 

Findings Observations 

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

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

issues (findings and observations) is provided in Attachment 5 of this report. In addition, 

Section 1.2 summarizes the findings. 

Table 1-1: Issue Matrix for the INL Site Profile 

Descriptions (a) 

Issue 

Classification 

Objective 1: 

Completeness 

Of Data 

Objective 2: 

Technical 

Accuracy 

Objective 3: 

Adequacy of 

Data 

Objective 4: 

Site Profile 

Consistency 

Objective 5: 

Regulatory 

Compliance (b) 

Issue 1: (5.1.1.1) Routine Airborne 

Releases Finding (5) X X X X Category 2 

Issue 2: (5.1.1.2) Episodic Airborne 

Release Finding (6) X X X X Category 2 

Issue 3: (5.1.1.3) Direct Gamma 

Exposures Finding (7) X X X X Category 2 

Issue 4: (5.1.2.1) Completeness and 

Quality of INL Internal Dosimetry 

Programs Finding (8) X X X X Category 3 

Issue 5: (5.1.2.2) High-Risk Jobs 

(Internal Exposure) Finding (9) X Category 1 

Issue 6: (5.1.2.3) Calibration of Internal 

Dosimetry Analytical and Monitoring Observation X X 

Equipment 

Issue 7: (5.1.2.4) Changes of Internal 

Dose Limits Observation X 

Issue 8: (5.1.2.5) High Fired Plutonium 

and Uranium Intakes Finding (10) X Category 1 

Issue 9: (5.1.2.6) Skin and Facial 

Contamination Observation X Category 3 

Issue 10: (5.1.2.7) Breathing Rates Observation X 

Issue 11: (5.1.2.8) Non-Occupational 

Worker Elimination of DU Background Finding (11) X X X Category 2 

Issue 12: (5.1.2.9) Unmonitored 

Workers Observation X 

Issue 13: (5.1.2.10) Naval Reactor 

Facility Workers Observation X Category 2 

Issue 14: (5.1.2.11) Plutonium 

Monitoring Observation X X X Category 1 

Issue 15: (5.1.3) SL-1 Accident Dose 

Reconstructions Finding (1) X X X 

Issue 16: (5.1.4.1.1) Completeness and 

Quality of INL Beta/Gamma Dosimetry Finding (8) X X X Category 3 

and Record Keeping Programs 

Issue 17: (5.1.4.1.2) Penetrating and 

Non-Penetrating Doses Finding (4) X X X X Category 3 

Issue 18: (5.1.4.1.3) Correction For 

Beta Doses Observation X X 

Issue 19: (5.1.4.1.4) Angular 

Dependence Correction Factor for Observation X 

Gamma Dose 
 

 

 

 

 

 

 

 

 

 

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Table 1-1: Issue Matrix for the INL Site Profile 

Descriptions (a) 

Issue 

Classification 

Objective 1: 

Completeness 

Of Data 

Objective 2: 

Technical 

Accuracy 

Objective 3: 

Adequacy of 

Data 

Objective 4: 

Site Profile 

Consistency 

Objective 5: 

Regulatory 

Compliance (b) 

Issue 20: (5.1.4.1.5) Restating Beta 

Dose As Gamma Dose Observation X 

Issue 21: (5.1.4.1.6) Photon Spectrum 

Split Observation X X 

Issue 22: (5.1.4.1.7) Immersion Dose Observation X X X 

Issue 23: (5.1.4.1.8) High-Risk Jobs 

(Beta/Gamma Exposure) Finding (9) X Category 1 

Issue 24: (5.1.4.1.9) Extremity Dose Observation X Category 2 

Issue 25: (5.1.4.1.10) Discrepancies 

between PIC and Film Reading Observation X 

Issue 26: (5.1.4.1.11) Minimum 

Detection Limit Observation X X 

Issue 27: (5.1.4.1.12) Minimum 

Reporting Level (Beta/Gamma) Finding (3) X X X Category 2 

Issue 28: (5.1.4.2.1) Minimum 

Reporting Level (Neutron) Finding (3) X X X Category 2 

Issue 29: (5.1.4.2.2) Failure to Properly 

Address Neutron Exposures Finding (2) X X X X Category 1 

Issue 30: (5.1.4.2.3) Neutron 

Calibration Deficiencies Finding (2) X X X Category 3 

Issue 31: (5.1.4.2.4) Completeness and 

Quality of INL Neutron Dosimetry and Finding (8) X X X X Category 3 

Record Keeping Programs 

Issue 32: (5.1.4.2.5) Uncertainty 

Estimation for Neutron Doses Observation X X X 

Issue 33: (5.1.4.2.6) Neutron Organ 

Dose Observation X 

Issue 34: (5.1.4.2.7) High-Risk Jobs 

(Neutron Exposure) Finding (9) X Category 1 

Issue 35: (5.1.4.2.8) Multiplying 

Factors for Missed Neutron Dose Observation X X X X Category 3 

Table Notes: 

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

(b) Category 1: Least challenged by any deficiencies in available dose/monitoring data are dose reconstructions for 

which even a partial assessment (or minimized dose(s)) corresponds to a probability of causation (POC) value in 

excess of 50%, and assures compensability to the claimant. 

Category 2: The use of upper bound values is limited to those instances where the resultant maximized doses 

yield POC values below 50%, which are not compensated. For this second category, the dose reconstructor 

needs only to ensure that all potential internal and external exposure pathways have been considered. 

Category 3: The most complex and challenging dose reconstruction represents claims where the case cannot be 

dealt with under one of the two other categories. 

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

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

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

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

within the report that addresses each issue. 
 

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

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

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

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

how the profile satisfies its intended purpose and scope. 

1.1 SUMMARY OF STRENGTHS 

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

internal, and external sources) based on historical operations, NIOSH compiled an enormous 

amount of data describing the radioactive materials, operations, and processes in the various 

facilities at INL. The associated TBDs generally provide a wealth of useful information to aid 

the dose reconstructors, who have the task of determining individual claimant radiation 

exposures. 

The INL Site Description TBD (Rohrig 2005), in addition to providing an overview of the site 

history, layout, facilities, and present status, and a framework in which to consider occupational 

exposures, is very thorough and detailed. It describes the past and current operations and 

missions of 14 areas and 101 facilities and processes. The TBD lists major radionuclides of 

concern at each facility and identifies potential sources of internal and external exposures. This 

information is very helpful to the dose reconstructions for the claimants. 

The Occupational Medical Dose TBD is thorough and methodical. For the purpose of estimating 

occupational medical exposure for workers at INL, NIOSH includes a table (Table 3A-1), which 

presents various organ doses resulting from chest x-ray examination for employees for the entire 

period of INL operation. This table is conveniently structured and easy to use. The organ doses 

were also reasonably estimated using available information on site practices and medical 

equipment. 

For airborne emissions from routine facility operations, the Occupational Environmental Dose 

TBD relies heavily on two previous works: (1) Idaho National Engineering Laboratory 

Historical Data Evaluation (INELHDE, DOE 1991a); and (2) Identification and Prioritization of 

Radionuclide Releases from the Idaho National Engineering and Environmental Laboratory, 

Final Report (RAC 2002). Using data from these two reports, NIOSH developed worker 

inhalation intake values for eight INL facilities (ANL, ARA, CFA, ICPP, RWMC, SPERT, TAN, 

TRA) for nine key radionuclides (Ce-144, I-131, Pm-147, Pu-238, Pu-239/240, Ru-106, Sr-89, 

Sr-90, and Y-91) from 1952 to 2002. These intake values are presented in Table 4-1 to Table 4-8 

of the TBD. For direct gamma exposures to environmental releases, NIOSH used facility fenceline TLD measurement data from Environmental Monitoring Data Reports (EMDRs) between 

1972 and 1983. The environmental gamma dose values are presented for eleven facilities in 

Table 4-13. For airborne emissions from episodic events, such as criticality accidents or special 

tests, NIOSH again used data from INELHDE and developed three tables providing worker 

inhalation intake values for primary radionuclides for criticality events (Table 4-10), special tests 

(Table 4-11), and initial engine tests (Table 4-12). These tables are helpful for dose 

reconstructions for certain claimants who worked at the INL site and, hence, were exposed to 

airborne emissions. 
 

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A prominent component of the INL Occupational Internal Dose TBD is Table 5.7-1, which 

presents default assumptions that could be used to calculate missed dose. Missed dose default 

assumptions are based on upper bound operating source terms from INL facilities. The table is 

easy and convenient for dose reconstructors to use if worker intake records are available and 

complete and exposures are not too high. On the other hand, this table does not include all 

potential sources of missed internal doses for workers; these will be discussed in the following 

sections. The table presents recommended missed dose default assumptions for four periods of 

operations as follows: 

(1) Beginning to 1960: The intake defaults are based on gross β activities in urine samples. 

A different, claimant-favorable weighting factor is applied to each of four radionuclides 

considered (Sr-90, Cs-137, Pu-238, and Ce-144). Default intake values are presented for 

personnel who worked at any of the INL facilities in this period, with an additional (in 

vivo) intake of I-131 added for personnel who worked at the TRA. 

(2) 1961 to 1970: The defaults are based on in-vivo (whole-body counting) measurement of 

Cs-137. A claimant-favorable weighting factor is applied to each radionuclide (Sr-90 and 

Pu-238). Default intake values are presented for personnel who worked at any of the INL 

facilities in this period, with additional intakes of Ce-144, I-131, Ag-110m, and Ta-182 

added for personnel who worked at the TRA. 

(3) 1971 to 1980: The defaults are based on in-vivo measurement of Cs-137. A claimantfavorable weighting factor is applied to each radionuclide (Sr-90 and Pu-238) for all 

facilities in this period. For those personnel who worked at the INTEC (previously ICPP) 

facility or an unknown INEL facility (e.g., NRF), an intake of Ce-144 is added. For those 

personnel who worked at ANL-W, intakes of Pu-239 and Ce-144 are added. For those 

personnel who worked at the TRA, an intake of I-131 is added. For those personnel who 

worked at other areas (other than INTEC, ANL-W, TRA, and unknown facilities), intakes 

of Pu-238 and Ce-144 are added. 

(4) 1981 to present: The defaults are based on in-vitro bioassay measurement. Intakes and 

weighting factors based on radionuclides associated with aluminum-clad fuel intakes are 

applied to personnel at all INL facilities except ANL-W, INTEC, and SMC in this period. 

For those who worked at ANL-W, intakes and weighting factors associated with stainless 

steel-clad fuel are used. For those who worked at INTEC or an unknown INL facility, 

intakes and weighting factors associated with zirconium-clad fuel are used. For those 

who worked at SMC, the DU uranium isotopic activities are used. 

Table 6B-1 pf the INL site occupational external exposure TBD presents maximum annual 

missed dose values for gamma radiation exposures for different films (e.g., 552 Dupont film, 

558 Dupont film, 508 Dupont film) and for different TLDs (LiF, LiF in Teflon, Harshaw TwoChip, Panasonic Four-Chip) based on the MRL (maximum recording level) for six different time 

periods; (1) 1951–1958, (2) 1958–1966, (3) 1966–1974, (4) 1974–1975, (5) 1974–1985, and 

(6) 1986 to present. The missed dose values are given for different dosimeter exchange 

frequencies (weekly, biweekly, monthly, quarterly, and semi-annually). For gamma dose 

reconstruction for a claimant, the equation (N x MRL/2) given in OCAS-IG-001 (Taulbee 2002) 
 

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should be used, where N is the number of zero dose results in a claimant’s file. There is no 

multiplying factor for gamma missed dose uncertainties. 

For missed neutron dose, Table 6B-2 of the external exposure TBD provides maximum annual 

missed dose values for different neutron dosimeter types (NTA film and TLD) based on MRL 

for three different time periods; (1) 1951–1958, (2) 1958–1976, and (3) 1976 to present. The 

table also shows the maximum annual missed neutron dose. For a neutron dose reconstruction 

for a claimant, the equation (N x MRL/2) given in OCAS-IG-001 (Taulbee 2002) should again 

be used; however, here N is the number of periods (week, month, or quarter) a worker had 

missed neutron dose. The TBD provides two multiplying factors when the MRL for NTA film is 

used in estimating the missed neutron dose. The missed neutron dose should be multiplied by 

1.25 for most workers and by 2 for workers on the MTR experimental floor for the period before 

October 1976. For the period after October 1976, no multiplying factor is needed. 

1.2 SUMMARY OF FINDINGS 

Table 1-1 presents a list of 35 issues that were identified during the review process. Seventeen 

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

findings. Those findings are summarized below in the order according to what SC&A believes is 

their level of significance to the dose reconstructions; references are given to report sections 

where more detailed discussions may be found. Note that 2 sets of 3 issues have been combined 

into Findings No. 8 and No. 9, and another 2 sets of 2 issues into Findings No. 2 and No. 3, 

reducing the final number of findings to 11. 

Finding 1 (Issue 15): SL-1 Accident Dose Reconstructions (Section 5.1.3) – The TBDs do 

not evaluate the potential missed internal and external doses or the associated uncertainties for 

the over 1,000 rescue and cleanup workers involved with the SL-1 accident that occurred in 

January 1961. There was a high potential for significant exposures, because the equipment used 

and the radiological control policies in place in that era were not as advanced and protective as 

those in current use. The TBDs should develop adjustment factors related to stay time, dose field 

estimates, internal dose results, external dose readings, and contamination level estimates. 

NIOSH did not evaluate potential missed neutron doses for the first responders and rescue 

workers of the accident. 

Finding 2 (Issues 29 and 30): Missed Neutron Dose (Sections 5.1.4.2.2 and 5.1.4.2.3) – The 

TBD presumes that neutron exposures at INL’s reactors are not a problem and, therefore, are 

adequately addressed. But INL had a total of 52 reactors most of which were experimental/ 

prototype in design, which typically operated with high-power densities and with minimum 

shielding and neutron moderation. It is, therefore, unjustified to presume that there are no 

missed neutron doses. In addition, there are deficiencies associated with neutron calibrations. 

Due to the use of the PoBe source for neutron calibration, dosimeters would significantly undermeasure neutron doses from sources with lower energy spectra. NIOSH should re-evaluate the 

entire approach in the TBD to account for potential missed neutron doses. 

Finding 3 (Issues 27 and 28): Minimum Reporting Levels (Sections 5.1.4.1.12 and 5.1.4.2.1) 

– NIOSH does not provide adequate information supporting the use of chosen detection 
 

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threshold levels to represent the MRL values for gamma film badges and TLDs. Most 

significantly, NIOSH’s approach for determining the MRL values for NTA emulsion film is not 

thorough and supported. For example, NIOSH uses 10 neutron readings in one data sheet from 

March 1958 to determine the MRL values for the period between 1951 and 1957, and 6 neutron 

readings to represent all neutron measurements between 1959 and 1976. Furthermore, the use of 

MRL/2 as the missed external dose for dose reconstruction per OCAS-IG-001 is not claimantfavorable for claims where the probability of causation value is close to 50%. In addition, 

NIOSH’s MRL values of 14 mrem and 20 mrem appear low and are inconsistent with generic 

values given for NTA dosimeters, as well as values cited by other DOE facilities with similar 

neutron source terms and detectors. NIOSH should re-evaluate the MRL values used and 

provide more supportable default values. 

Finding 4 (Issue 17): Penetrating and Non-Penetrating Doses (Section 5.1.4.1.2) – The 

procedures and algorithms used in the film badge dosimetry service in the early days 

underestimated the Hp(10) dose, because the low-energy photons reaching the dosimeter were 

considered beta radiation. Surprisingly, the film service then added this beta dose (to the skin) to 

the "deep" dose; a practice that is claimant favorable. However, the TBD also correctly requires 

the dose reconstructor to consider only the "deep dose" as Hp(10), but in doing so, the lowenergy photon contribution to Hp(10) is lost. To be claimant favorable, INL calculated the beta 

dose and the gamma dose, summed these two together, and recorded the result in the worker files 

as a whole-body dose. The current dose reconstruction process is applying the gamma dose 

correctly as the effective dose. The problem is that this gamma dose is not claimant favorable, as 

the information on dose due to low-energy gammas (E < 100 keV) has been lost. NIOSH should 

re-evaluate the missed gamma dose due to the deficiencies in the procedures and algorithms. 

Finding 5 (Issue 1): Routine Airborne Releases (Section 5.1.1.1) – For airborne emissions 

from routine facility operations, NIOSH relies heavily on two previous works: (1) Idaho 

National Engineering Laboratory Historical Data Evaluation (INELHDE, DOE 1991a); and (2) 

Identification and Prioritization of Radionuclide Releases from the Idaho National Engineering 

and Environmental Laboratory, Final Report (RAC 2002). SC&A found that the source terms 

provided require improvement for use in determining the worker intake from airborne releases at 

different INL facilities. The data NIOSH uses do not take into account the deficiencies in the 

environmental monitoring equipment and their locations, and, in addition, NIOSH does not 

assess the uncertainties associated with the meteorological dispersion model used for the INL 

site. Most importantly, the source terms do not account for worker inhalation of resuspended 

contaminated soils and materials around the INL facilities. 

Finding 6 (Issue 2): Episodic Airborne Releases (Section 5.1.1.2) – For airborne releases 

from episodic events (such as criticality and special tests), NIOSH again relies on the two 

primary documents, INELHDE and RAC 2002, for determining onsite concentrations of 

radionuclides. In a previous study (SC&A 2003), SC&A determined that the airborne releases 

associated with several of the Initial Engine Tests of the Aircraft Nuclear Propulsion (ANP) 

Program were likely to have been underestimated by factors ranging from 2 to 7. Also, NIOSH 

did not evaluate the uncertainties associated with the deficiencies in air monitoring equipment. 
 

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Finding 7 (Issue 3): Direct Gamma Exposures (Section 5.1.1.3) – For direct gamma exposure 

from environmental releases, NIOSH used fence-line TLD measurement data from 

Environmental Monitoring Data Reports (EMDRs) between 1972 and 1983. SC&A believes that 

these TLD measurements are not adequate for reconstructing direct gamma doses to personnel 

working outdoors at and around a specific INL facility inside the fence-line boundary. The 

NIOSH approach assumes all outdoor workers at a facility would receive an average direct 

gamma dose from a normalized ground concentration of radionuclides. If the assumption were 

valid, the fence-line TLD results should be multiplied by a weighting factor to account for 

uncertainties in TLD sensitivity and geometry. NIOSH has not done that in the TBD. However, 

NIOSH’s assumptions are not claimant favorable, because they do not take into account the most 

bounding scenarios; (1) personnel working outdoors may become immersed in the plume of 

routine or episodic releases from the facility stack, (2) personnel working outdoors may inhale 

resuspended cumulative ground radionuclide depositions or windblown contaminated soils from 

any neighboring dry ponds, and (3) the cumulative ground concentrations inside the fence line 

are generally higher than those near the fence line. The fence-line TLDs are too far from the 

bounding source terms to represent the actual direct gamma doses received by the outdoor 

workers. Therefore, this TBD approach does not appear to be claimant favorable. In addition, 

NIOSH should re-examine the direct gamma dose values, because some of these values 

presented in Table 4-13 of the TBD appear to be much lower than the recent INL-published 

values. 

Finding 8 (Issues 4, 16 and 31): Completeness and Quality of INL Radiological Protection, 

Personnel Dosimetry, and Record Keeping Programs (Sections 5.1.2.1, 5.1.4.1.1, and 

5.1.4.2.4) – The tone of the TBDs strongly suggests that NIOSH (i.e., the TBD authors) had full 

confidence in the radiological protection programs, the internal dosimetry programs, and the 

dosimetry record keeping systems at the INL site, in the past and at the present. The 

identification and determination of missed internal and external dose for workers are heavily 

influenced by this assumption of confidence, but SC&A found this premise to be unsupported 

after examining several critical DOE-HQ Tiger Team and DNFSB site audit reports. In addition, 

many site experts interviewed by SC&A indicated that there were significant deficiencies and 

inconsistencies in radiation work practices throughout the operating history of the INL facilities. 

These observations jeopardize the validity of the TBD approaches in reconstructing missed 

worker internal and external doses. 

Finding 9 (Issues 5, 23 and 34): High-Risk Jobs (Sections 5.1.2.2, 5.1.4.8, and 5.1.4.2.7) – 

Site experts interviewed by SC&A classified INL as an “acute dose” site, with a significant 

number of facilities, operations, experiments, and occurrences providing the possibility of 

personnel receiving dangerous levels of radiation. However, NIOSH did not evaluate 

comprehensively the facility and field data to identify and separate out the high-risk or high-dose 

jobs for worker internal and external exposures. For example, the TBD does not evaluate the 

potential missed external dose (extremity, bladder, lens, or prostate) received by a pipefitter, who 

worked on a reactor vessel or process waste tank in a tight space environment where the 

dosimeter measurements would not be effective and representative due to angular dependence or 

bodily shielding. This information is essential for dose reconstructors to fill in the data gap when 

dose records in a claimant’s file are not complete. 
 

 

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Finding 10 (Issue 8): High-Fired Plutonium and Uranium Intakes (Section 5.1.2.5) – The 

Occupational Internal Dose TBD did not evaluate the hazard associated with high-fired 

plutonium and uranium at the INTEC (ICPP) and RWMC facilities. High fired Pu-238, Pu-239, 

and uranium are not easily dissolvable, nor do they readily break into very small particles. They 

also emit some gamma rays and neutrons. Similar to the treatment of recycled uranium, NIOSH 

should evaluate the lung dose for intake of high-fired uranium and plutonium oxide particulates 

(alveolar deposition). 

Finding 11 (Issue 11): Non-Occupational Worker Elimination of DU Background (Section 

5.1.2.8) – The derivation of the background value of 0.16 μg/L used for subtraction from each 

urinalysis result of uranium prior to assessment of occupational internal dose for SMC radiation 

workers is not technically sound. The baseline background (population) intake value was 

determined by a study of urine samples submitted by non-radiation workers at the SMC facility. 

A better approach would be to use the urine excretion samples by non-INL people in the Idaho 

Falls areas. This approach would not create a suspected bias due to uranium intake through 

various pathways (inhalation and ingestion) by non-radiation workers while working at the SMC 

facility. During the site expert interview, the dosimetry staff indicated that they tried to use 

residents from the Idaho Falls area, but none were willing to sign a liability waiver form. In a 

subsequent study, the dosimetry staff used sixteen non-radiation workers from the INL CFA 

facility. In addition, this background value is much higher than the national survey data. 

NIOSH should consider this subtraction from urinalysis results as a missed internal dose. 

1.3 OPPORTUNITIES FOR IMPROVEMENT 

Not withstanding the many positive, helpful features of the TBDs in providing guidance to the 

dose reconstructors, SC&A identified a number of areas that represent potential opportunities for 

improvement. These are summarized below: 

Incomplete Dose Records in Worker Files: NIOSH should look into the possibility of many 

missing dose records in worker files. From interviews of retired, past, and current workers, and 

also from audit reports (DOE-HQ and DNFSB), there appear to be many incident reports, 

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

the worker records. Most of these reports were kept, historically, at the INL facilities. NIOSH 

holds the position that the dosimetry records reflect accurately all the doses workers received 

while working at INL. The suspected incompleteness of the worker records is a serious issue, 

since it may lead to significant underestimation of workers’ radiation dose. 

Use of Untrained Workers: NIOSH should look into the worker utilization practices at INL in 

the early years (from 1950s to 1980s). From interviews of retired, past, and current workers, it 

appears that INL often sent workers, even untrained ones (e.g., secretaries, yardmen, etc.), from 

one INL facility to perform emergency or high-dose radiation work in another INL facility. 

After these workers reached their weekly or bi-weekly dose limit, in most cases 200 mrem, they 

would be sent back to their original workplace. For example, INTEC (ICPP) is one of the 

facilities that required frequent utilization of workers from other INL facilities to augment the 

work force, due to its high contamination and high radiation level (acute dose) environment. 

Several site experts interviewed indicated that they personally had the experience of being sent to 
 

 

 

 

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INTEC to be “burned” to their dose limit. In addition, some workers interviewed indicated that 

they had experienced that the doses they received after being “burned” at the other facilities were 

frequently not recorded in their worker exposure files. 

Angular Dependence Correction Factor for External Dosimeter: NIOSH should provide 

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

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

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

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

tissues/organs below the waist. 

Breathing Rate: In the INL occupational environmental dose evaluation (TBD Section 4.2.1), 

NIOSH uses an at-rest breathing rate, which is not claimant favorable.3 

Plutonium Monitoring: NIOSH should provide historical information on the plutonium 

analysis methods used at the INL site. It is entirely possible that selective plutonium monitoring 

of workers was used at INL until 1980. Without this information, the dose reconstructors would 

not be able to assign missed internal dose, due to plutonium intakes in the time period before 

1980. 

Dose Calculation Example: NIOSH should provide an example (or examples) in the TBD text 

of a hypothetical dose reconstruction using recorded records, missed dose assignment, and dose 

assignments when dosimeters read zero dose. 

Consistent Air Dispersion Model: NIOSH should use a consistent methodology for calculation 

of occupational dose that is appropriate for application to onsite workers. In addition, the 

components of environmental dose should be consistent among the DOE facilities evaluated. 

For routine releases, the INL Occupational Environmental Dose TBD uses the MESODIF model, 

employing an objective regional trajectory computational scheme combined with the Gaussian 

diffusion equation for a continuous point source. It is a forward time-marching Gaussian plume 

model in which successive, small plume elements (or puffs) are advected throughout the 

computational area. For episodic releases, the TBD used the RSAC program primarily for 

calculating onsite meteorological dispersion parameters for the various airborne release incidents. 

RSAC provides the option to use various types of meteorological diffusion models and 

parameters applicable at INL, all of which are based on the Gaussian plume model. In the case 

of Hanford Site and SRS, for example, a RATCHET puff model and the Gaussian model were 

used, respectively. 

Recycled Uranium and Plutonium: The dose contribution from trace radionuclides in recycled 

uranium and plutonium should be evaluated in terms of dose to particular organs of concern and 

the relative impact on internal dose reconstruction. NIOSH should evaluate the lack of formal 

policies for considering trace radionuclides in recycled uranium and plutonium, and develop 

3

 See Attachment 1, Occupational Environmental Dose, Question/Response No. 7, of this report for further 

discussion of the at-rest breathing rate assumption. 
 

 

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bounding conditions that can be applied at all applicable DOE facilities, including INTEC 

(formerly ICPP) at the INL site. 

Beta/Gamma Dosimeter Adjustments and Uncertainties: A method to consistently account 

for laboratory, radiological, and environmental uncertainties in dosimeter readings should be 

developed and appropriately applied to recorded dosimeter results, so that it is clear what sigma 

value should be entered into Interactive RadioEpidemiological Program (IREP) Parameter 2. 

Tank Farm Worker: NIOSH should complete an evaluation of the relative hazards associated 

with work at the Tank Farms and the completeness of monitoring related to Tank Farm workers, 

including subcontractor and construction workers. 

Use of Site Expert Input: NIOSH should make a greater effort to take into account site expert 

information and to investigate worker accounts. First-hand experience and association with the 

INL facilities enable site experts and workers to provide original perspectives and information 

concerning site practices and exposure histories that may not appear in the official records. 

Missed Dose and Off-normal Practices: NIOSH should evaluate the significance of offnormal practices for missed dose by analysis of film badge data and site expert interviews. This 

is essential to determine if there were areas or periods where badges may not have been 

consistently worn when the actual dose was near the administrative control limit. 

Data Uncertainty: The statutory requirement of a claimant-favorable dose reconstruction 

process is achieved by (1) giving the benefit of doubt when there are unknowns, and (2) defining 

uncertainties for measured data and selecting the 99th percentile values of a Monte Carlo 

distribution. In the site profile TBDs, it is often found that lower percentile values are used 

instead of the 99th percentile values as default assumptions for missed dose calculations. NIOSH 

should re-examine all the data uncertainty values used in the TBDs to ensure that this statutory 

requirement of claimant-favorable dose reconstruction is met. 
 

 

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2.0 SCOPE AND INTRODUCTION 

INL is an 890-square-mile reservation encompassing almost 572,000 acres with a 

maximum distance of about 39 miles from north to south and 36 miles from east 

to west. It is 30 to 60 miles west of Idaho Falls, Idaho. Major site-related 

research facilities and offices are in Idaho Falls. The site, situated on the Snake 

River Plain of southeastern Idaho at an elevation of about 5,000 ft., is above the 

Snake River Plain Aquifer. In 1949, the U.S. Atomic Energy Commission (AEC) 

established the National Reactor Testing Station (NRTS) in Idaho as a Federal 

reservation to build, test, and operate nuclear reactors. The site utilized a variety 

of support facilities and equipment. In 1974, the NRTS became the Idaho 

National Engineering Laboratory (INEL) and, in 1997, the Idaho National 

Engineering and Environmental Laboratory (INEEL). On February 1, 2005, the 

site became the Idaho National Laboratory (INL) combining the research side of 

the INEEL and ANL-W and the Idaho Cleanup Project (ICP) working on closure 

of inactive portions of the site. INL is unique among U.S. Department of Energy 

(DOE) facilities in its size and complexity and diversity of it facilities and 

operations, with many independent technical areas, contractors, goals, and 

missions. (Rohrig 2005, pg. 13) 

Under the Energy Employees Occupational Illness Compensation Program Act of 2000 

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

Reconstruction Under the Energy Employees Occupational Illness Compensation Program, of 

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

Advisory Board on Radiation and Worker Health (Advisory Board) is mandated to conduct an 

independent review of the methods and procedures used by the National Institute for 

Occupational Safety and Health (NIOSH) and its contractors for dose reconstruction. As a 

contractor to the Advisory Board, S. Cohen and Associates (SC&A) has been charged to support 

the Advisory Board in this effort by independently evaluating a select number of site profiles that 

correspond to specific facilities at which energy employees worked and were exposed to ionizing 

radiation. 

This report provides a review of the following Technical Basis Documents (TBDs) related to 

historical occupational exposures at INL and which, together, constitute the site profile: 

• 	 ORAUT-TKBS-0007-1, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Introduction (Rohrig 2004i) 

• 	 ORAUT-TKBS-0007-2, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Site Description (Rohrig 2005) 

• 	 ORAUT-TKBS-0007-3, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational Medical Dose (Rohrig 2004m) 
 

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• 	 ORAUT-TKBS-0007-4, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) –Occupational Environmental Dose (Peterson 

2004) 

• 	 ORAUT-TKBS-0007-5, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational Internal Dose (Rich and Wenzel 

2004) 

• 	 ORAUT-TKBS-0007-6, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational External Dosimetry (Rohrig 

2004e) 

SC&A has critically evaluated the INL TBDs and other applicable documents in order to 

accomplish the following: 

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

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

individual dose reconstructions 

• 	 Assess the technical merit of the data/information 

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

SC&A’s review of the six TBDs focuses on the quality and completeness of the data 

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

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

the objectives stated in SC&A’s Standard Operating Procedure for Performing Site Profile 

Reviews (SC&A 2004a). The criteria for evaluation include whether the TBDs provide a basis 

for scientifically supportable dose reconstruction in a manner that is adequate, complete, efficient, 

and claimant favorable. Specifically, these criteria were viewed from the prospective of whether 

dose reconstructions based on the TBD would support robust compensation decisions. 

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

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

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

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

reconstructions, bridging uncertainties, or correcting technical inaccuracies. 

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

workers were exposed and, supplementing the workers’ individual dose records, determine 

claimant exposures through different pathways over time. The hierarchy of data used for 

developing dose reconstruction methodologies is (1) dosimeter readings and bioassay data, 

(2) co-worker data and workplace monitoring data, and (3) process description information or 

source term data. 

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

procedure, this report is organized into the following sections: 
 

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(1) Executive Summary 

(2) Scope and Introduction 

(3) Assessment Criteria and Method 

(4) Site Profile Strengths 

(5) Vertical Issues 

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

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

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

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

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

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

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

addressed, would further improve the TBDs and may possibly reveal deficiencies that will need 

to be addressed in future revisions of the TBDs. In this review, SC&A has identified 35 issues, 

categorized into 11 findings (some issues are combined), and 18 observations. The TBDs, in 

many ways, have done a successful job in addressing a series of technical challenges. In other 

areas, the TBDs exhibit shortcomings that may influence some dose reconstructions in a 

substantial manner. 

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

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

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

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

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

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

how the profile satisfies its intended purpose and scope. 
 

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3.0 ASSESSMENT CRITERIA AND METHOD 

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

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

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

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

3.1 OBJECTIVES 

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

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

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

These objectives are discussed in the following sections: 

3.1.1 Objective 1: Completeness of Data Sources 

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

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

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

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

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

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

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

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

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

sources in the TBDs, was evaluated to determine the relevance and use of the data collected by 

NIOSH to the development of the site profile. Additionally, SC&A evaluated selected records 

publicly available relating to the INL site and records provided by site experts. SC&A requested 

a number of documents during the site expert interview visit, which have all been provided by 

INL. 

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 INL. The goal of this objective is to first analyze the data according to sound 

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

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

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

3.1.3 Objective 3: Adequacy of Data 

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

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

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

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

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

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

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

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

3.1.4 Objective 4: Consistency Among Site Profiles 

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

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

accomplish this objective, the INL TBDs were compared to the Hanford and Savannah River 

Site (SRS) TBDs. Both the Hanford and SRS site profiles are appropriate for comparison as the 

sites, all large DOE laboratories, had similar missions. Attachment 4 consists of a table 

comparing key assumptions and approaches taken by site profiles for the three sites. This 

assessment was conducted to identify areas of inconsistencies and determine the potential 

significance of any inconsistencies with regard to the dose reconstruction process. 

3.1.5 Objective 5: Regulatory Compliance 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

produces a POC of greater than or equal to 50%. 
 

 

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Category 2: A second category of dose reconstruction is defined by Federal guidance, which 

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

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

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

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

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

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

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

and external exposure pathways have been considered. 

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

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

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

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

to compensate a claim. 

Although simplistic in design, to satisfy this type of a dose reconstruction, the TBD must, at a 

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

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

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

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

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

concentrations and surface contaminations must be identified. 

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

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

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

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

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

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

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

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

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

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

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

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

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

the doubt to claimants. 

3.2 ASSESSMENT METHODOLOGY 

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

reviewed each of the six TBDs, their supplemental attachments, TIBs, and other relevant 

documents, giving due consideration to the three categories of dose reconstructions that the site 

profile is intended to support (Section 3.1.5). The TBDs generally provide well-organized and 
 

 

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somewhat or partially user-friendly information for the dose reconstructor when adequate data 

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

of the fact that the site profile is not required by the EEOICPA or by 42 CFR Part 82, which 

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

reconstructors for identifying site-specific practices, parameter values, and factors that are 

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

employment and exposure record. Based on information provided by NIOSH personnel, SC&A 

understands that site profiles are living documents, which are revised, refined, and supplemented 

with TIBs as required, to help dose reconstructors. Site profiles are not intended to be 

prescriptive nor necessarily complete in terms of addressing every possible issue that may be 

relevant to a given dose reconstruction. In addition, they are not intended for the “layman,” but 

for the health physics personnel immersed in the review process. The principal documents and 

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

informing SC&A’s assessment, are the following: 

• 	 ORAUT-TKBS-0007-1, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Introduction (Rohrig 2004i), explains the 

purpose and the scope of the site profile. SC&A was attentive to this section, because it 

provides a useful overview and explains the role of each TBD in support of the dose 

reconstruction process. Hence, the introduction helps in framing the scope of the site 

profile. As will be discussed later in this report, NIOSH may want to include 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 site profile. 

NIOSH may also want to include a roadmap to the dose reconstructor expounding on the 

entire dose reconstruction process and showing how the TBDs fit into that process. 

• 	 ORAUT-TKBS-0007-2, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Site Description (Rohrig 2005), is an 

extremely important document, because it provides a description of the facilities and 

processes, as well as historical information that serve as the underpinning for subsequent 

INL TBDs. Specifically, this document describes the history and current status of 14 

areas, 101 facilities and processes, and their associated source terms that are relevant to 

dose reconstruction. SC&A’s review of this section addresses whether all the potentially 

important site activities and processes are described, and whether characterization of 

source terms is complete and sufficient to support dose reconstruction. 

• 	 ORAUT-TKBS-0007-3, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL)– Occupational Medical Dose (Rohrig 2004m), 

provides a set of procedures and default dose values for reconstructing the radiation 

exposures of workers from medical radiographic procedures that were required of 

employees at INL. SC&A reviewed this section for completeness and technical adequacy. 

• 	 ORAUT-TKBS-0007-4, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL)– Occupational Environmental Dose (Peterson 

2004), provides background information and guidance to dose reconstructors for 

reconstructing the doses to unmonitored workers outside of the facilities at INL who may 
 

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have been exposed to routine and episodic airborne emissions from these facilities. 

SC&A reviewed this section from the perspective of the source terms and the 

atmospheric transport, deposition, and resuspension models used to derive the external 

and internal doses to these workers. In addition, SC&A also reviewed the Supplement to 

Technical Basis Document 4 for the Idaho National Engineering and Environmental 

Laboratory: INEEL Occupational Environmental Dose (Peterson 2004s). 

• 	 ORAUT-TKBS-0007-5, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational Internal Dose (Rich and Wenzel 

2004), presents background information and guidance to dose reconstructors for deriving 

occupational internal doses to workers. This section was reviewed with respect to 

information and guidance regarding the types, mixes, and chemical forms of the 

radionuclides that may have been inhaled or ingested by the workers, the recommended 

assumptions for use in reconstructing internal doses based on whole-body counts and 

bioassay data, the methods recommended for use in the reconstruction of missed internal 

doses, and the methods recommended for characterizing uncertainty in the reconstructed 

internal doses. 

• 	 ORAUT-TKBS-0007-6, Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational External Dose (Rohrig 2004e), 

presents background information and guidance to dose reconstructors for deriving 

occupational external doses to workers. This section was reviewed with respect to 

information and guidance regarding the different types of external radiation (i.e., gamma, 

beta, and neutron) and the energy distribution of this radiation to which the workers may 

have been exposed. SC&A also reviewed the recommendations for converting external 

dosimetry data to organ-specific doses, the methods recommended for reconstructing 

missed external doses, and the methods recommended for characterizing uncertainty in 

the reconstructed external doses. 

• 	 ORAUT-OTIB-0009, Technical Information Bulletin in Support of INEEL Technical 

Basis Document Section 6: Reanalysis of Hankins MTR Bonner Sphere (Rohrig 2004h), 

presents re-analyzed neutron dose equivalents of the original Hankins measured values 

for three IREP energy groups at the MTR floor (Hankins 1961). This TIB also provides a 

distribution of the neutron dose equivalent among the energy groups and their 

corresponding fractional NTA film responses. 

• 	 Supplement to Technical Basis Document 4 for the INEEL: INEEL Occupational 

Environmental Dose (Peterson 2004s) presents supportive material for occupational 

environmental dose information and data. In particular, Attachment 4-A of this 

supplement provides 18 tables listing 14 total organ doses (thyroid, skin, lungs, bone, 

breast, stomach, ULI, LLI, red marrow, gonads, SI, spleen, liver, kidneys) for the 

inhalation and air-immersion pathways at different INL facilities, due to either 

operational releases or episodic events from other facilities. The reason why these organ 

dose values are not included in the TBD is not specified; but the TBD indicates that the 

air-immersion dose calculations are not necessary, because NIOSH believes that they 

would be recorded in the fence-line TLD doses (pg. 24). 
 

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

assessment of the six TBDs, their supporting documentation, and several TIBs. SC&A then 

submitted questions to NIOSH with regard to assumptions and methodologies used in the site 

profile. A teleconference was conducted with staff members of NIOSH, ORAU, an ORAU 

subcontractor (Intrepid), and the SC&A team to discuss SC&A’s questions, and to allow NIOSH 

to provide explanations and clarifications. NIOSH, ORAU, and ORAU subcontractor personnel 

subsequently were given the opportunity to comment on a draft summary of the teleconference. 

A summary of the questions, conference call, and some subsequent clarifications is provided in 

Attachment 1 of this report. 

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

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

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

especially when trying to recall information from decades ago, the interviews, especially taken in 

the aggregate, provided much useful insight from the perspective of the INL workers themselves. 

Attachment 3 provides summaries of the interviews conducted by SC&A in the Idaho Falls area 

during the course of this review. The site experts interviewed include current, former, and 

retired staff from dosimetry, radiation control, operations, environmental monitoring, 

maintenance, instrumentation, electrical, mechanical, security, engineering, management, and 

other support organizations. Each summary is an edited paraphrase of conversations held with a 

number of site experts, rather than a verbatim transcript. Personnel statements have been 

grouped into categories to provide a linkage with various portions of the INL Site Profile. 

References that may identify specific site experts have been omitted for privacy reasons. 

Interviewed individuals were given the opportunity to review the interview summary for 

accuracy. This is an important safeguard against missing key issues or misinterpreting some 

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

comments on the summaries. The DOE-ID Classification Office also reviewed the summaries. 

Information provided in the teleconference with NIOSH mentioned above was evaluated against 

the preliminary findings to finalize the vertical issues addressed in the report (Section 5). There 

are three levels of review for this report. First, SC&A team members review the report internally. 

Second, SC&A engages an outside consultant, who has not participated in the preparation of this 

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

cycle, will consist of a review of this draft by the Advisory Board and NIOSH. The first two 

review levels have been completed. 

After the Advisory Board and NIOSH (and its contractors and subcontractors) have an 

opportunity to review this draft, SC&A plans to request a meeting with Advisory Board 

members and NIOSH representatives to discuss the report. Following this meeting, SC&A will 

revise the draft and deliver the final version to the Advisory Board and to NIOSH. We anticipate 

that, in accordance with the procedures followed during previous site profile reviews, the report 

will then be published on the NIOSH web site and discussed at an Advisory Board meeting. 

This last step in the review cycle completes SC&A’s role in the review process, unless the 

Advisory Board requests SC&A to participate in additional discussions regarding the closeout of 

issues, or if NIOSH issues revisions to the TBDs or additional TIBs, and the Advisory Board 

requests SC&A to review these documents. 
 

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Finally, it is important to note that SC&A’s review of the TBDs and their supporting TIBs is not 

exhaustive. These are large, complex documents, and SC&A used its judgment in selecting 

those issues that we believe would be important with respect to dose reconstruction. 
 

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4.0 SITE PROFILE STRENGTHS 

In developing a TBD, the assumptions used must be fair, consistent, and scientifically robust, 

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

development of the TBD must also consider efficiency in the process of analyzing individual 

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

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

in the INL TBDs. These strengths are described in the following sections. 

4.1 COMPLETENESS OF DATA SOURCES 

The INL TBDs exhibit the following strengths in terms of the completeness of their data 

sources: 

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

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

and historical records, the most complete and expansive of which is Susan Stacy’s book, 

commissioned by DOE to commemorate 50 years of INL operations, Proving the 

Principle, A History of the Idaho National Engineering and Environmental Laboratory, 

1949–1999 (Stacy 2000). In developing the site profile, NIOSH drew upon information 

contained in 287 reports and documents cited in the reference sections of five of the six 

TBDs. Facilities are grouped into 14 areas and 101 facilities and processes. A concerted 

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

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

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

compilation to be an important strength of the report. 

(2) NIOSH made an admirable effort to compile historical data on medical x-ray equipment 

and techniques used at INL. From these data, NIOSH was able to develop a 

comprehensive, user-friendly table listing organ doses from occupational medical x-ray 

exposure to workers. For missing data prior to 1954, NIOSH has been able to use 

estimates from a document prepared by Ron Kathren (Kathren 2003) to fill the gap. 

Notwithstanding SC&A’s opinion that there are opportunities for improvement in treating 

the uncertainties associated with the derivation of the organ doses, SC&A considers the 

medical exposure section to be one of the strengths of the site profile. 

(3) In compiling the atmospheric source terms for deriving outdoor occupational exposures 

to unmonitored and monitored workers, NIOSH depended heavily on two reports; 

INELHDE (DOE 1991a) and RAC 2002. In addition, a supporting document, 

Supplement to Technical Basis Document 4 for the Idaho National Engineering and 

Environmental Laboratory: INEEL Occupational Environmental Dose (Peterson 2004), 

was developed to provide more detailed information on the data and methods used in 

deriving the potential worker intake values at 8 primary INL facilities for routine releases, 

potential worker intake values for identified episodic events, and direct gamma exposure 

values at 11 INL facilities. Though these reports are comprehensive and provide much 

useful information and guidance, SC&A believes that there are opportunities for 
 

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improvement in identifying other (missing) source terms and the methods used to 

reconstruct the missed environmental gamma doses. 

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

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

radioactive materials at the various INL facilities, and describing the relevant operations 

and their associated processes. Notwithstanding this achievement, there are opportunities 

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

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

due to deficient work practices and inadequate instrumentation. 

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

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

external radiation sources at various INL facilities, and describing the relevant operations 

and their associated processes. NIOSH also compiled data related to external dosimetry 

used at INL over its entire operating history. From these gathered data, NIOSH 

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

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

dose values and identifying high-risk jobs. 

4.2 TECHNICAL ACCURACY/CLAIMANT-FAVORABILITY 

The INL site TBDs exhibit the following strengths, in terms of their technical accuracy and 

claimant-favorability: 

(1) Medical x-ray procedures have been investigated thoroughly to determine the 

radiographic techniques used at INL. The frequencies of medical examinations 

performed on new hires and on workers over the operating years have been exhaustively 

reviewed and determined. The INL Site Occupational Medical Dose TBD (Rohrig 

2004m) provides methods for calculating doses from both posterior-anterior (PA; back to 

front) and lateral chest x-rays in Section 3.4 (pg. 6). For dose calculations, NIOSH 

provides default skin entrance air kerma values for PA and lateral chest x-ray 

examinations. It was determined that no photofluoroscopic examination has ever been 

performed at the INL site. Table 3A-1 of the TBD presents the recommended default 

organ doses for six time periods of INL operations. An assigned uncertainty of ±30% at 

1 sigma (84% confidence) was recommended for the calculated dose results. 

(2) NIOSH made a concerted effort to determine the airborne release source terms due to 

routine operations from major INL facilities and episodic events, such as unplanned 

criticalities and planned special fuel tests. NIOSH also thoroughly evaluated the 

environmental monitoring data over the operating history to determine the fence-line 

dose results. The Occupational Environmental Dose TBD provides a set of tables listing 

the worker intake quantities by year due to routine operations (Tables 4-1 to 4-8) and 

episodic events (Table 4-10 to 4-11), and the direct gamma dose to workers (Table 4-12) 

at primary INL facilities. The overall accuracy of the information provided is judged 

acceptable, however, NIOSH did not include other potential sources of missed dose to 
 

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workers, such as resuspended contaminated soils and accumulated ground depositions. 

In addition, NIOSH did not evaluate the validity and quality of the monitoring data to 

determine uncertainties associated with the source terms. 

(3) NIOSH identified the primary contributing facilities to worker internal exposures by 

extensively analyzing the operations and processes at different INL facilities. NIOSH 

also evaluated the radiologically significant radionuclides for these operations and 

processes. The most limiting source terms for worker internal exposure were then 

determined to arise from the processed nuclear fuels at ICPP (Table 5.6.2.5-1, pg. 32). 

The TBD selected the most limiting radionuclides in the fuel inventory as the default 

worker intakes to calculate the missed internal doses at different INL facilities by 

multiplying by a set of weighting factors (Table 5.7-1, pg. 38). However, NIOSH did not 

evaluate other potential sources of missed internal dose to workers, which may have been 

due to inconsistent and deficient field work practices. In addition, NIOSH did not 

identify potential high-risk or high-internal dose jobs leading to missed worker internal 

doses. 

(4) NIOSH made a concerted effort to determine the minimum reporting levels (MRLs) that 

were associated with the various types of dosimeters used over the life of the lab, for the 

different types of operations, and for the different types of external exposures (Rohrig 

2004e, Tables 6B-1 and 6B-2, pg. 47). In addition, NIOSH determined the potential 

gamma, beta, and neutron radiation fields at different INL facilities (Section 6.3.4). The 

assessment was not as complete and as accurate as it could have been, however, as 

NIOSH did not evaluate other potential sources of missed external dose to workers due to 

inconsistent and deficient field work practices. NIOSH also did not identify potential 

high-risk or high-external dose jobs, which may have led to significant missed worker 

external doses. 

(5)	 Similar to the Hanford site profile, the development of separate TBDs for the six primary 

areas, i.e., introduction, site description, Occupational Medical Dose, occupational 

environmental dose, occupational internal dose, and occupational external dosimetry, 

forms a model for later site TBDs. The format of the INL TBDs is more user-friendly 

than some of the previous TBDs, such as for Savannah River. The use of examples is 

beneficial to the dose reconstructor, as they summarize and clarify the suggested 

procedures to be followed. NIOSH should expand the presentation of examples in 

revised or new site profile TBDs; for instance, there is only one example provided in the 

Occupational External Dose TBD. 

4.3 ADEQUACY OF DATA 

The developers of the TBDs benefited from having access to information and data that were 

compiled as a part of several INL site programs. The following summarizes these programs and 

how the TBDs benefit from them, and also points out areas where the programs may have been 

deficient or where the TBDs could have made better use of the available data. 
 

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(1) While INL’s radiological control program has undergone continuous improvement since 

the inception of site operations, there are significant concerns about deficiencies in the 

program and its implementation in the early years of the site history. Radiological 

control personnel have implemented upgraded procedures and technology over time to 

reduce radiation dose to workers, and have enhanced personnel monitoring programs as 

more advanced equipment and instrumentation were added. 

(2) The personnel monitoring program utilized timekeeping, film badges (later TLDs), 

pocket ionization chambers (PICs), and air sampling to monitor and control worker 

exposure. The two-element film badge was used as early as August 1951 (Rohrig 2004e), 

but INL continued to use timekeeping and PICs to monitor real-time personnel dose after 

the implementation of the film badge. However, these PIC data were not provided in the 

TBD for dose reconstructors to use, although there were significant concerns about the 

validity of the film badge results among many site experts interviewed. 

(3)	 The INL dosimetry program initially used neutron film badges with Kodak nuclear track 

emulsion Type A (NTA) for monitoring worker exposure to neutron fields. Because of 

the missed dose from any neutrons below the NTA energy threshold of 0.5 to 0.8 MeV, 

INL switched to Hankins albedo dosimeters (Rohrig 2004e). 

(4) The external dosimetry staff characterized the workplace radiation fields at INL in 

response to a Tiger Team finding, making field measurements with a NaI(Tl) 

spectrometer and TLDs mounted on a phantom. The measurement results were 

compared to determine relative bias. These data, however, were not used in determining 

the missed external gamma and neutron doses. 

(5) In 1996, the INL external dosimetry staff performed neutron radiation field 

characterization to determine a facility neutron correction factor (FNCF) from the 9-inch 

to 3-inch ratio in the worker location. This FNCF is used to adjust the measurement 

result to dose equivalent. 

(6) INL has identified the MTR experimental floor as the largest contributor of neutron 

exposures to workers. In 1961, INL used 2-, 3-, and 8-inch polyethylene Bonner balls in 

a cadmium shield to characterize the intermediate and fast neutron at 21 locations around 

the MTR floor. Measurements of thermal neutron components were also made at six 

other locations. These 1961 data are reanalyzed in the TBD (Rohrig 2004h) using more 

recent Bonner response curves. Correction factors for missed neutron dose are then 

estimated. 

(7) Routine bioassay involving in-vitro and urine sampling were implemented in the 

beginning of the site history (probably from 1953). Whole-body counting became 

routine in the 1960s. However, thyroid counting used for high-risk workers was not 

mentioned in the TBD. The frequencies of the bioassay program evolved from an annual 

routine program to an event-based program in the later years, but there are significant 

gaps in these data, which need to be addressed. NIOSH gives incomplete instructions on 

how to use bioassay data to calculate doses. 
 

 

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(8) INL implemented environmental monitoring from the beginning of its operational history. 

All of the airborne releases and their impact to the populations around INL boundaries 

were reviewed and analyzed. With the help of NOAA, hourly meteorological data were 

available for the analyses. All the releases and analyzed results up to 1991 are 

documented in the INLEHDE (DOE 1991a). 

4.4 CONSISTENCY AMONG THE SITE PROFILES 

Although INL, Hanford, and Savannah River Site (SRS) missions overlapped to a significant 

extent, there are many differences in the facility designs and processes, and radiological practices. 

Attachment 4 compares many parameters and assumptions of these three sites. In some cases, 

these differences require site-specific assumptions in dose determinations. For example, due to 

many different nuclear reactor experiments and fuel reprocessing operations at INL, there is a 

long list of radionuclides of concern, including Cs-137, Co-60, Sr-90, Pu-239, U-235, and I-131. 

INL, however, did not have to consider the substantial particulate releases of ruthenium from the 

REDOX facility at Hanford; nor did SRS, whose facilities were built later. NIOSH has made a 

concerted effort to recognize and address site-specific issues in the TBDs. With respect to the 

Interactive RadioEpidemiologic Program (IREP) input parameters, the INL, Hanford, and SRS 

TBDs are consistent in many cases. This consistency was especially apparent in the 

occupational medical exposure sections, as seen in Table A.4-1 in Attachment 4. 

4.5 REGULATORY COMPLIANCE 

The TBDs’ reliance on environmental monitoring data, but only limited personnel data, to 

determine dose is not entirely consistent with the requirements outlined in 42 CFR Part 82, 

which specifies the hierarchy of data that is to be used in dose reconstruction. NIOSH has not 

effectively complied with the hierarchy of data required in §82.2 and its implementation guides 

for monitored workers. The two examples that follow show that NIOSH did not compile 

pertinent data, like bioassays and external doses, and dose rates. 

(1) Where in-vivo and in-vitro analyses are available, this information is provided for use in 

the determination of internal dose. However, the Occupational Internal Dose TBD relies 

entirely on reprocessed fuel information as the bounding source terms for potential 

worker inhalation intake. 

(2) The Occupational External Dose TBD does not use beta/gamma and neutron dosimetry 

data in the determination of external exposure. NIOSH uses the MRL values for different 

types of film badges and dosimeters to calculate the potential maximum external (gamma 

and neutron) dose to workers. Where environmental measurements are available, these 

data are used as the basis for environmental dose. 
 

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5.0 VERTICAL ISSUES 

SC&A has developed a list of issues regarding the INL site profile, which relate to the five 

objectives defined in the SC&A site profile review procedure (SC&A 2004a). Some issues 

pertain to a particular objective, while others to several objectives. A matrix relating the 

objectives and the relative importance of each issue is provided in Section 6.0. The issues 

identified as findings map into the four broad categories discussed in Section 5.1, and the issues 

identified as observations map into the two broad categories discussed in Section 5.2. Many of 

the issues raised are applicable to other DOE and Atomic Weapons Employer sites, and should 

be considered in the preparation and revision of other site profiles. 

5.1 DISCUSSION OF ISSUES 

5.1.1 Occupational Environmental Dose Issues 

The Occupational Environmental Dose TBD divides potential occupational environmental 

exposure to workers at the INL facilities into two pathways; (1) airborne releases from INL 

facilities, and (2) direct gamma doses resulting from facility operations. The airborne releases 

are further subdivided into routine operational releases and episodic releases. NIOSH used the 

source terms in the document, Idaho National Engineering Laboratory Historical Dose 

Evaluation Report (INELHDE), to develop a set of annual inhaled quantities (Bq/yr) for routine 

operational releases for 8 INL facilities for the dose reconstructors to choose from. NIOSH also 

developed intakes (Bq/event) for all identified episodic releases in the history of INL (from 1952 

to 2002). For direct radiation to workers, NIOSH used exclusively fence-line environmental 

TLD records to develop direct gamma values for 11 INL facilities from 1952 to 2002. 

5.1.1.1 Routine Airborne Releases 

For airborne emissions from routine facility operations, the Occupational Environmental Dose 

TBD relies heavily on two previous work; the INELHDE Report (DOE 1991a) and Identification 

and Prioritization of Radionuclide Releases from the Idaho National Engineering and 

Environmental Laboratory, Final Report (RAC 2002). The INELHDE project was chartered in 

December 1988 in response to inquiries concerning possible radiological consequences to the 

public from past projects and facility operations at INL. The purpose of the report is to provide 

estimates of potential offsite radiation doses encompassing the entire operating history of INL 

using a consistent methodology for all years. While potential ground water, surface water, and 

biotic pathways exist, they are not significant contributors to the offsite radiation dose, because 

the principal route by which radioactivity released on the INL site can reach offsite locations is 

through the airborne pathway. Consequently, this report concentrates on estimates of doses 

resulting from radioactivity releases to the atmosphere that are subsequently transported offsite 

through airborne pathways. 

Atmospheric releases are categorized as either operational or episodic. Operational releases are 

continuous and fairly uniform releases, occurring over a year or over a portion of a year, that 

span a variety of meteorological conditions; the report uses annual average meteorological 

conditions. Episodic releases took place over a short period of time, typically a few hours, and 
 

 

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may have been associated with a particular planned or unplanned event, such as the SL-1 

accident. Some tests as long as several months were modeled as episodic releases. Episodic 

releases are treated as distinct events using the actual meteorological conditions measured at the 

time of each release. The primary source of information for estimating the operational releases 

of radioactivity to the atmosphere is the measurement of stack effluents, as reported in the 

Radioactive Waste Management Information System (RWMIS) database. The report performed 

supplemental calculations using information from other technical reports to determine the 

radionuclide composition of the effluent. For episodic releases, estimates are taken from 

technical reports on individual INL facilities and test programs. 

The total amount of radioactivity associated with annual operational releases at INL was greatest 

in the late 1950s and early 1960s, and has decreased since then. The peak release of 

radioactivity in the 1980s was approximately one-tenth of that released in 1961. A total of 108 

individual episodic releases are considered in the INELHDE, and detailed dose calculations were 

done for the 54 most important releases. Most of the important episodic releases were associated 

with reactor fuel development tests conducted before 1961 (DOE 1991a, pp. v–vi). 

In the INELHDE document, calculation of atmospheric transport to locations off the INL site 

was done using the MESODIF computer code. The name MESODIF is derived from the terms 

mesocale (MESO) and diffusion (DIF) to indicate that the dispersion predictions of the model are 

valid for distances up to about 150 km (95 miles) from the release point (DOE 1991a, pg. 15). 

Data for dispersion calculations were taken from the records of the INL network of 

meteorological stations. Even though the INELHDE was developed only for evaluating the 

offsite population dose impact due to INL releases, the exposure pathways are similar for the 

onsite workers. Radiological assessment calculations were made for four exposure pathways; 

(1) external exposure from immersion in contaminated air, (2) external exposure from 

radioactivity deposited on the ground, (3) internal exposure from inhalation of contaminated air, 

and (4) internal exposure from ingestion of contaminated agricultural products. However, the 

INL Environmental TBD focuses on radiological assessment calculations for pathway (3). It 

does not evaluate pathway (1), and inadequately evaluates pathway (2). As for pathway (4), the 

TBD does not consider any ingestion scenarios. 

The TBD indicates that potential offsite doses from INL activities have been small. The largest 

radiation doses were calculated for an infant in 1956, when the effective dose equivalent (EDE) 

from operational and episodic releases was estimated to be 61 mrem and the thyroid dose 

equivalent was estimated to be 1,350 mrem. The trend in later years has been toward smaller 

doses as total INL site releases have declined. Episodic releases have made a substantial 

contribution to the total potential radiation dose only for a few years during the test program of 

the 1950s and 1960s. Radiation doses from airborne releases over much of the operating history 

of INL were small, compared to doses from natural background radiation (DOE 1991a, pg. v). 

To determine onsite concentrations of radionuclides from operational releases at INL facilities, 

NIOSH used the same methodology employed in the INELHDE. The release source term for 

each year of operation is the same as that documented in the INELHDE, with some minor 

modifications (see discussion below). With one exception, NIOSH performed an analysis to 

reduce the number of radionuclides from 56 to 9 for operational releases that contributed less 
 

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than 95% of the inhalation doses. For operational releases, the TBD used annual average 

meteorological conditions (mesoscale dispersion isopleths of ground-level air concentrations) 

from Bowman 1984 for the years prior to 1993. For 1993 to 2002, NIOSH used average 

meteorological conditions from annual environmental reports (ESRF 1994, 1995, 1996, 1997, 

1998, 2000; Stoller 2002a, 2002b, and 2002c) (TBD, pg. 10). For routine releases, NIOSH 

developed worker inhalation-intake values for 8 INL facilities (ANL-W, ARA, CFA, ICPP, 

RWMC, SPERT, TAN, TRA) for 9 key radionuclides (Ce-144, I-131, Pm-147, Pu-238, Pu239/240, Ru-106, Sr-89, Sr-90, and Y-91) from 1952 to 2002. These intake values (Bq/yr) are 

presented in Table 4-1 to Table 4-8 of the Occupational Environmental Dose TBD (Peterson 

2004). The following quotations are from that TBD and set the stage for the discussions that 

follow in subsequent sections. 

Other than the early releases from the GE-ANP IET operations, which are treated 

as episodic releases, by far the greater majority of the airborne radiological 

releases were contributed by the ICPP, the TRA, and EBR-I reactor, in that order. 

The ICPP airborne releases peaked in 1959 at about 1.3 million curies, dropped 

to about 800,000 curies the year after, rose to about 1.2 million the next year and 

then declined steadily to less than 200,000 curies in 1964 and remained at that 

level until 1975 when it rose for a year to about 250,000 curies. Since 1976 the 

annual releases from ICPP have been less than 200,000 curies. The TRA annual 

releases peaked at slightly less than one half of the ICPP releases (about 600,000 

curies) in 1963 and have declined to less than 200,000 curies in 1967 and have 

remained less than 100,000 curies since 1969. The ANL-W airborne releases, 

generally associated with EBR-II, peaked at 19,000 curies in 1965 and dropped 

steadily to less than 1000 curies in 1969 where it has remained since. 

Annual site boundary doses for the maximally exposed individual reported in the 

Annual Environmental Monitoring Reports assume the total INEEL airborne 

effluent is released from a 250 ft. (76 m) stack midpoint between TRA and ICPP. 

Meteorological dispersion for the year is calculated for ground-level elevations 

by NOAA for the release height as presented in the average mesoscale dispersion 

isopleths. Originally, in the INEEL 4 TBD, INEEL facility personnel intakes for 

operational releases (Bq/yr) were calculated using INEEL total annual release 

radionuclides and respective quantities documented in RAC 2002. These release 

radionuclides and quantities are the result of a critical review by RAC of the 

releases documented in the INELHDE, performed in 1990 and 1991. The 

INELHDE took its operational release source term from the Radioactive Waste 

Management Information Service (RWMIS), the official documentation for 

gaseous, liquid, and solid radioactive wastes produced by facilities at the INEEL. 

However, when the INELHDE evaluation was performed, some minor problems 

were identified with respect to the annual source terms attributed to the TRA and 

ICPP releases as documented in the RWMIS. Identification and correction of 

these problems are discussed in length in pages A-12 through A-44 of the 

INELHDE. In the late 1990s, the State of Idaho requested the Centers for Disease 

Control (CDC) to critically review the INELHDE, which was contracted out to 

the RAC. The RAC review of the INELHDE “improved” the source terms for the 
 

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annual operational releases as described in pages 10 through 23 of RAC 2002. In 

summary, the changes to the RWMIS annual releases made for the INELHDE 

were in some cases substantial, but annual curie amounts remained unchanged. 

The changes in the INELHDE source terms made by the RAC were more for 

completeness and were not substantial. All changes preserved the annual total 

curies that are documented in RWMIS. In the final RAC assessment, the list for 

annual releases included 56 radionuclides, which was preserved when the first 

version of the INEEL 4 TBD was written. 

The formal review of INEEL 4 TBD requested that the number of radionuclides be 

reduced from 56 such that 95% of the original dose was preserved. The reduction 

was accomplished by using an option provided by the RSAC-6 code that allowed 

the total CEDE to be listed by radionuclide in ascending dose order. The 

reanalysis using this option allowed the number of radionuclides to be reduced 

from 56 to 16. After that reduction was accomplished, a further review suggested 

that the number of radionuclides be reduced to 10 and still preserve 95% of the 

original dose. The list was reduced to 11 radionuclides using the ICRP 26/30 

methodology with the help of the RSAC-6 code. Calculations at that time showed 

that eliminating one more radionuclide would have reduced the original total 

CEDE to 92%–93% of the original dose. The final review suggested that the final 

list include only those nuclides contained in a list treated by the Integrated 

Modules for Bioassay Analysis (IMBA), which meant that the two radionuclides, 

132I and 89Rb, which contributed the least dose of the 11 be omitted. The final 

reduction, completed with a table of doses, created by the ICRP 60/66 

methodology for the INEEL operational releases, indicated that 132I and 89Rb 

each contributed only about 0.7% of the semi-final dose. Therefore, those two 

radionuclides were deleted leaving the total list of radionuclides to a total of nine, 

all of which are treated by the IMBA code used by the dose reconstructors. 

(Peterson 2004, pp. 9–10) 

5.1.1.1.1 Completeness and Quality of Release Data Used 

(1) The INELHDE report uses effluent release data primarily from stack monitoring and air 

sampling systems installed across the INL site. There were a total of 23 air samplers; 

12 within the site, and 11 outside the INL boundaries. The INELHDE claims: 

The best and most comprehensive available data on releases of 

radioactivity to the atmosphere from INEL facilities have been used. The 

primary source of information on operational releases was the 

measurement of stack effluents as reported in the Radioactive Waste 

Management Information System Information System (RWMIS) data base. 

The data in RWMIS are generally most complete and reliable for the years 

since this data base was established in 1971. Data for the years from 

1962 to 1970 are less complete and reflect less sophisticated monitoring 

instrumentation and record keeping systems. Data from 1952 to 1961 do 

not identify the specific mixture of radionuclides released; therefore 
 

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additional assumptions and calculations were required to determine the 

radionuclide composition of the effluent and perform the dose assessment 

(DOE 1991a, pg. 11). 

The ambient air monitors at various INL site locations had been found to be deficient in 

meeting the siting requirements specified in 40 CFR 58 (near obstructions) and meeting 

minimum flow rate for particulates specified in 40 CFR 50 (e.g., 2.5 cfm versus 39 cfm), 

as cited in a DOE report, Tiger Team Assessment of the Idaho National Engineering 

Laboratory, DOE/EH-0178, 1991 (DOE-HQ 1991). NIOSH should evaluate the 

adequacy of the stack release data for use in the TBD. The Tiger Team report states the 

following: 

Deficiencies exist within data residing in the RWMIS data base in that the 

data is not subject to formal validation and verification procedures by the 

generator before the annual certification is made. RESL does not believe 

that verification/validation is part of their responsibility; however, no 

other entity at INEL is performing that function. In one instance, data has 

been omitted from this data and not identified through any 

verification/validation process. The example noted by the Tiger Team 

involved the omission of Krypton-85m, Krypton-87, and Krypton-88 from 

the emission inventory reports for the ZPPR at ANL-W between 1980 and 

1991. (Vol. 1, pg. 3-265) 

There is one particulate radiation monitor in the immediate vicinity of 

ICPP. This monitor is located near the ICPP fence line but is not 

downwind of ICPP radiation sources during predominant wind directions, 

and therefore is not sited properly to observe most low-level radiation 

releases from ICPP. (Vol.1, pg. 3-48) 

The ambient air monitors observed did not meet the siting requirements 

specified in 40 CFR 58, Appendix E, Paragraph 8.2, with respect to 

locating monitors near obstructions (i.e., buildings, parking lots) and 

maintaining an unrestricted airflow 270 degrees around the air sampler. 

For example, the ambient air monitor located at RESL is within 10 feet of 

the corner of CF-676 behind CF-690, and the ambient air monitors 

located at TAN, and ANL-W are located in parking lots adjacent to the 

fenceline. (Vol. 1, pg. 3-50) 

The type of ambient air monitors being used by RESL and Rockwell are of 

the low volume variety. The flow rate of the air samplers vary between 1.5 

and 2.5 cubic feet per minute. 40 CFR 50, Appendix B, states that the 

minimum and maximum flow rate for particulate monitors is 39 to 60 

cubic feet per minute, respectively. (Vol. 1, pg. 3-50) 

The height of the ambient air monitors maintained by RESL and Rockwell 

is approximately 1 meter. 40 CFR 58, Appendix E, states that the required 
 

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height of the air intake for large scale samplers is 2–15 meters. (Vol. 1, 

pg. 3-50) 

Radiological effluent sampling and monitoring systems throughout INL facilities had not 

been evaluated to ensure that they would detect, quantify, and respond adequately to 

unplanned releases (DOE-HQ 1991). Also, airborne effluent particulates from INL 

operations had not been adequately characterized, and measurements and sampling 

techniques did not ensure a representative sample for effluent monitoring systems (DOEHQ 1991). Several quotations from the Tiger Team report illustrate these deficiencies 

(and point to areas that NIOSH should investigate further): 

A review of the radiological effluent sampling and monitoring systems at 

contactor facilities throughout the INEL has indicated that the majority of 

the stacks with the potential for release of radioactive particulates in the 

effluent have not been evaluated to ensure that particulate sampling is 

representative or that it will detect, quantify, and adequately respond to 

unplanned releases of radioactive material to the environment as required 

in DOE 5400.5. ANSI N13.1-1969 provides data which shows that the 

longer the sample delivery line and the greater the number of 90 degree 

turns, the greater the potential for particulate losses (depending on 

particulate sizes, sampling flow rate, particulate density, etc.). Examples 

of facilities in question are described below. (Vol. 1, pg. 3-59) 

The MTR Stack (TRA-710), the Hot Cell (TRA-632), Science & 

Technology Building (TRA-604), alpha Wing Addition (TRA-661), and the 

ATR stack (TRA-771) have not been evaluated for particulate losses in the 

sample delivery lines, sampling probe integrity, and the proper placement 

of the sampling probe. Sample delivery lines are typically too long and 

possess 90 degree bends. A study has not been conducted to determine 

whether or not the sample collection flow is isokinetic. An EG&G internal 

review of the ATR stack stated that ‘there is no evidence that 

representative samples are taken.’ (Vol. 1, pg. 3-59) 

Flow rate measurements are not taken for the TRA-604 stack sampling 

system as required by DOE 5400.1, Chapter IV, Section 5.a.(2). The 

Sampling system flow rate is determined by the power of the motor 

providing the suction. This method also fails to correct for flow rate 

changes due to filter loading or equipment malfunctions. (Vol. 1, pg. 3-59) 

The CFA laundry fails to take flow rate measurements at the beginning 

and end of each sampling period to account for dust loading (I-R-204). 

Additionally, during air sample filter analysis, the wrong alpha selfabsorption factor was being used per EG&G Idaho procedures (R-223). 

These factors will cause an underestimation of the effluent release 

concentrations. (Vol. 1, pg. 3-59) 
 

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The air sample extraction systems on the stacks at PBF-WERF Incinerator 

and TAN Hot Cell/Hot Shop have not been evaluated for particulate losses 

resulting in the sample delivery line (I-R-257, I-R-274). (Vol. 1, pg. 3-59) 

The TAN Decontamination Shop Stack was found to have no stack 

monitoring system after the high efficiency particulate air (HEPA) 

filtration. This facility is not presently utilized. (Vol. 1, pg. 3-59) 

The RWMC Drum Vent System stack monitoring system has not been 

evaluated for particulate losses in the sample delivery line. The sample 

delivery line possesses three 90 degree bends, and is approximately 

20 feet long. This facility is currently not in use and a larger facility is 

expected to house the operation, however, without any planned upgrades 

to the stack monitoring system other than extending the stack height. 

EG&G Idaho has determined that this facility will be required to perform 

continuous monitoring under NESHAP when operational and , hence, the 

stack monitoring system will be required to conform to 40 CFR 61.93 and 

ANSI N13.1-1969 (I-R-427). It currently does not meet these standards. 

(Vol. 1, pp. 3-59 and 3-60) 

The Main Stack and FAST stack continuous, on-line radiological 

monitoring systems for airborne effluents collect samples of the 

particulate in the stack gas through the use of a slip-stream. The slipstream is drawn at a constant flow from the primary sample delivery line, 

and particulates are collected on a bulk filter. The slip-stream sample 

delivery line extends away from the primary sample delivery line at a 45 

degree angle incident to the direction of the flow. This primary sample 

delivery line has been characterized for particulate losses during routine 

operations and unplanned release events (R-477). However, the 

particulate losses and the adequacy of the slip-stream design to respond in 

a timely manner during an unplanned release event has not been 

determined. (Vol. 1, pg. 3-61) 

The stack monitoring delivery systems at the NWCF have not been 

evaluated for particulate losses in the sample delivery lines from the 

calciner and the HVAC system and have not been evaluated to determine 

whether the sample flow rate is isokinetic. The sample delivery lines 

exceed 30 feet in length and have at least three 90 degree bends. (Vol. 1, 

pg. 3–61) 

ICPP identified 49 sources that were evaluated for radioactive air source 

emissions as part of the document prepared for the NESHAP waiver. This 

document presents the release rates prior to abatement for sources of 

radioactive air emissions in worksheet form. Some of the abated release 

rates provided by ICPP are not consistent with what is reported in 

previous Environmental Monitoring Reports. For example, the NESHAP 
 

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emission report presents the FAST stack unabated release for tritium as 

2.31 curies, which does not reflect operating conditions. The 1987 and 

1988 Annual Monitoring Reports present the measured releases for 

tritium as 480 and 183 curies, respectively. .. Approximately 150 air 

radioactive emission sources were presented in INEL’s Air Emission 

Inventory. As a result of INEL’s review of source terms, only three 

sources of radioactive emissions were only slightly below the limit 

requiring continuous monitoring (e.g., 0.067 mrem/yr from Vent 008 at 

CPP-627) and were rejected as requiring continuous monitoring 

(continuous monitoring required at 0.1 mrem/yr). Considering the above 

deficiencies in source term characterization, the potential exists that other 

sources of radioactive air emissions at INEL may require continuous 

monitoring as prescribed in 40 CFR 61.93 in addition to those already 

identified. (Vol. 1, pg. 3-67) 

Given these deficiencies in the INL stack monitoring and air sampling systems identified 

by the DOE Tiger Team audit, it is unlikely that the INELHDE results would be complete 

and representative of the actual effluent releases from different INL facilities. This 

would further impact the quality and the validity of the dose assessments. NIOSH should 

evaluate the uncertainties associated with these issues, so that the recommended worker 

intake values from environmental releases would be truly claimant favorable. In 2003, 

SC&A performed a study of radioactive release source terms for two major INL 

programs for the CDC (SC&A 2003). This study reviewed stack monitoring data and 

uncertainties associated with the data. SC&A provided this report to NIOSH for use in 

the preparation of the INL Site Profile. 

(2) During the site expert interview conducted by SC&A, INL environmental staff indicated 

that other unplanned (episodic) releases occurred at different facilities that are not 

included in the INELHDE. Therefore, they are also not included in the NIOSH TBD. 

For instance, there was an incident at the INTEC (ICPP) in the early 1990s where 

particulate releases were observed as a result of a new steam cleaning process of the CPP 

stack. The airborne material released was believed to be Cs-137 attached to white 

insulation material. Measurable radioactivity was associated with these releases. This 

information is not included in the TBD for the use of dose reconstructions. 

5.1.1.1.2 Dispersion Model 

The Environmental TBD uses a mesoscale model (MESODIF), which employs an objective 

regional trajectory computational scheme, combined with the Gaussian diffusion equation for a 

continuous point source, to estimate dispersion for transport of releases. It is a forward timemarching Gaussian plume model in which successive, small plume elements (or puffs) are 

advected throughout the computational area. The following quotations, taken from the 

INELHDE (DOE 1991a), describe the use of the mesoscale dispersion data from MESODIF 

model for the calculations of average ground level air concentrations at INL facilities: 
 

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Since 1973, meteorological information gathered from the 26-station INEL 

telemetry network in the Upper Snake River Plain has been evaluated annually 

using the MESODIF computer code. The results were illustrated in dispersion 

isopleth maps contained in annual INEL Site environmental monitoring reports. 

Adequate meteorological information was not available before 1973 or in 1978, 

when the telemetry system was being upgraded, to produce annual dispersion 

isopleth maps. For these years, the MESODIF computer code was run on a 9­

year (1974–1983, but excluding 1978) set of meteorological data from the 

telemetry system. The resulting “average dispersion conditions” were used in 

computing the annual impact from operational releases before 1973 and in 1978. 

The annual dispersion isopleth maps prepared from the MESODIF computer code 

output were used from 1973 through 1989, except for 1978. The annual average 

wind speed toward the inhabited location of highest annual average 

concentration from a point midway between the Idaho Chemical Processing Plant 

(ICPP) and Test Reactor Area (TRA) was estimated based on the joint frequency 

distribution of wind speeds. The wind speed distribution for 9 years (1974–1983) 

was used to estimate the annual average wind speed before 1984. The annual 

distributions were used for this purpose beginning in 1984. The detailed data 

from a grid of 176 points, on which the dispersion isopleth map is based, were 

used to calculate the ground-level dispersion coefficient at Atomic City, the 

location of the maximally exposed individual for the 9-year average. This was 

done with a log-log interpolation from the four grid points surrounding Atomic 

City. (DOE 1991a, Appendix B, pg. B-9) 

MESODIF was also used to calculate dispersion coefficients for episodic events 

occurring at the INEL. The data flow for discrete episodic events is shown in 

Figure B-i. For these individual events MESODIF was run on a finer scale grid, 

grid point spacing of 2 mi as opposed to 5 1/3 mi for the annual operations. 

Available meteorological data for the years before 1967 were limited to wind 

sensor tower sites located at the Central Facilities Area (CFA) and Initial Engine 

Test (JET). The accuracy of the plume trajectories calculated for the past events 

depends on how well these two wind stations were able to represent the actual 

winds affecting the plume. It is reasonable to assume that the further the plume 

traveled from the wind tower locations, the less reliable were the calculations. 

Most of the events were initiated within a few miles of one of these sites, so the 

important initial portions of these calculated trajectories are considered 

reasonable representations. (DOE 1991a, Appendix B, pg. B-11) 

MESODIF is a trajectory model that specifically requires spatial information describing upper 

boundary layer meteorological conditions. INL, however, does not have a real-time database 

that could be used in defining model trajectories. This caused significant uncertainties in 

accurately estimating the dispersion of released materials. This deficiency was noted in the 

DOE-HQ 1991 Tiger Team report (DOE-HQ 1991), a contemporaneous report of the INELHDE: 

The site has an adequate surface meteorological observation network, but does 

not conduct regular monitoring to describe meteorological conditions above the 
 

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heights of the surface meteorological towers, and does not have a complete upper 

air database. The INEL site encompasses 890 square miles and has only three 

meteorological towers with observation levels above 50 feet, two of which have 

observation levels at 200 feet and one at 250 feet. The remainder of the boundary 

layer, above 250 feel to approximately 5,000 feet, is not characterized. The INEL 

has numerous sources which utilize stacks equal to or greater than 200 feet to 

vent air emissions. Plume rise from these sources was frequently observed during 

this assessment to exceed 500 feet above the ground, and the air emissions are 

transported in an atmospheric region that that site is currently not 

characterizing.... Upper air data is available from the National Weather Service 

offices in Boise, ID and Salt Lake City, UT, but these data are not representative 

of the INEL site conditions given the close proximity and orientation of the 

mountains to the site. (DOE-HQ 1991, Vol. 1, pp. 3-56 and 3-57) 

RESL performs an analysis of all onsite monitor locations using the results 

generated by the mesoscale diffusion (MESODIF) model. The MESODIF model 

is designed for transport distances of greater than 20 km. This model is not 

appropriate for evaluating monitor locations within 20 km, and this type of model 

is not recommended by EPA-540/4-87-007 for evaluating appropriateness of 

monitor locations within these transport distances. (DOE-HQ 1991, Vol. 1, 

pg. 3-47) 

RESL has performed an evaluation of offsite monitors using the annual average 

results of the annual MESODIF and CAP-88 National Emission Standards of 

Hazardous Air Pollutants evaluation. As discussed in Finding A/CF-6, the lack of 

a representative upper air meteorological data for the INEL limits the accuracy of 

the MESODIF model because it requires upper air data to define representative 

trajectories for plumes. Without this data, it cannot be verified that offsite (>20 

km) monitors are located to be representative of maximum INEL related pollutant 

concentrations. (DOE-HQ 1991, Vol. 1, pg. 3-47) 

First, the mesoscale model used by INL in the INELHDE is only appropriate for evaluating 

dispersion coefficients for locations at greater than 20 km distance, and is not appropriate for 

those facilities that are within 20 km of each other. It is definitely not suitable for determining 

dispersion of airborne releases within several hundred feet from a facility building or a stack. 

Many INL facilities, however, are within 20 km from each other. Second, even for facilities 

located more than 20 km from each other, the dispersion coefficients calculated by this model are 

deficient and not representative. Third, and most important, the workers considered in this site 

profile are more impacted by the release plumes at the facility where they worked than those 

from more distant facilities; this mesoscale model is not capable of addressing such short 

distance dispersion coefficient factors. NIOSH should re-examine the validity of the mesoscale 

model data used in the occupational environmental TBD. 
 

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5.1.1.1.3 Other Observations 

NIOSH should list the routine airborne release activities and associated uncertainties for each 

INL facility in the Occupational Environmental Dose TBD. This data would be helpful for the 

dose reconstructors to assess whether the worker intakes are applicable to the claim they are 

considering. An example would be useful, showing how the worker exposure could be 

calculated using the release activities, uncertainty values, and weighting factors. 

5.1.1.2 Episodic Airborne Releases 

To determine onsite concentrations of radionuclides from episodic releases at INL facilities, 

NIOSH used the Radiological Safety Analysis Computer code, RSAC-6 (Wenzel and Schrader 

2001; Peterson 2004s) for the pathway calculations. NIOSH employed the total list of 

radionuclides that are applicable to the respective episodic release from INELHDE. In the 

Occupational Environmental Dose TBD, NIOSH developed three tables providing worker 

inhalation intake values (Bq/event) for important radionuclides for (1) seven criticality events 

(Table 4-10), (2) special tests (Table 4-11), and (3) initial engine tests (Table 4-12). Peterson, in 

the environmental TBD supplement, wrote about the use of RSAC: 

For the INEEL technical basis document (TBD), the RSAC program was used 

primarily for calculating on-site meteorological dispersion parameters for the 

various airborne release incidents; radiological doses due to these airborne 

releases was a minor additional calculation and was originally to be 

incorporated into the TBD only as an indication of the magnitude of the dose 

since they were relatively small and easy to calculate. These doses were 

originally incorporated into the TBD to be an indicator to the dose reconstructor 

only of the magnitude of the relative doses. They were removed as a result of 

reviewer comments. (Peterson 2004s, pg. 7) 

The RSAC program has the ability to calculate a fission product inventory for a 

given reactor operation (or a criticality) that is comparable to an ORIGEN fission 

product inventory calculation, the industry standard. The program can then 

tailor modeled releases to emulate actual releases from the standpoint of material 

released, attenuate the material by filters or by holdup within a volume 

containment, and for the rate of release, either instantaneously or over a period 

of time. The program provides the option to use various types of meteorological 

diffusion and contains the meteorological parameters applicable at the INEEL, 

all of which are based on the Gaussian plume model. From the meteorologically 

dispersed material, the program calculates, according to the ICRP 26 and ICRP 

30 methodology, radiological doses from the ingestion, inhalation, air-immersion, 

cloud-gamma, and deposition pathways to 19 different organs of the body and a 

corresponding external effective dose equivalent (EDE) or a committed effective 

dose equivalent (CEDE) to the whole body, depending on the pathway for dose 

calculation. All of the doses discussed here were calculated with default 

parameters (plume centerline, particle size, solubility, etc.) such that radiological 

doses are maximized. (Peterson 2004s, pg. 7) 
 

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The code has been extensively used for safety analysis work at the INEEL since 

1965 when it was used for dose calculations for the Safety Analysis Report for the 

Advanced Test Reactor. After the publication of the INELHDE (DOE 1991[a]), 

the Centers for Disease Control (CDC) and the State of Idaho instigated a 

technical review of the methodology for the analyses for this report. As a result of 

this technical review, minor changes were recommended in some of the airborne 

source terms. These minor changes did not change the total curies released, but 

added small amounts of other radionuclides and respective quantities to earlier 

years where detection had not been as ‘low level’ as in the more recent years. In 

the course of the review, the Radiological Assessment Corporation (RAC) 

examined the RSAC-4 program that defined the radiological doses for each of the 

releases. (Peterson 2004s, pg. 8) 

SC&A considers use of the RSAC code and underlying model adequate for calculating 

dispersion factors and the resulting dose to workers. 

5.1.1.2.1 Completeness and Quality of Episodic Releases Data 

The airborne releases associated with several of the Initial Engine Tests (IETs 3, 4, and 10) of 

the Aircraft Nuclear Propulsion (ANP) Program, as estimated by the INELHDE, were likely to 

have been underestimated as follows: 

• 	 IET 3 – underestimate of total radionuclide release by up to a factor of about 3 

• 	 IET 4 – underestimate of noble gases by up to a factor of about 16, halogens by up to a 

factor of about 7, and solids by a factor of up to about 2 

• 	 IET 10 – underestimate of total radionuclide releases by up to a factor of about 7 

These concerns were cited in the SC&A report, A Critical Review of Source Term for Select 

Initial Engine Tests Associated with the Aircraft Nuclear Propulsion Program at INEL (SC&A 

2003, -g. 2-24), which states the following: 

The HDE Task Group acknowledged the absence of available raw effluent data as 

well as the deficiencies/limitations of summary data contained in the report by 

Thornton et al. (1962b). The HDE, therefore, modeled release estimates that 

were principally based on historical operating records and photographic 

evidence, which characterized the extent of fuel damage to the HTRE No. 1 

reactor core. ... Embedded in the HDE model of radioactive releases are several 

assumptions that potentially may have underestimated the true release quantities 

of fission products. Identified below are four key model parameters whose values 

may have differed significantly from those assumed by the HDE. 
 

 

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5.1.1.2.2 Other Observations 

Similar to the problem with routine releases, NIOSH should list the episodic airborne release 

activities for each INL facility used in the Occupational Environmental Dose TBD. 

Uncertainties associated with the release activities should also be provided. This data would be 

helpful for the dose reconstructors to assess whether the worker intakes are applicable to their 

claim. An example may be given showing how the worker exposure could be calculated using 

the release activities, uncertainty values, and weighting factors. 

5.1.1.3 Direct Gamma Exposure 

For direct gamma exposures to environmental releases, NIOSH used facility fence-line TLD 

measurement data from Environmental Monitoring Data Reports (EMDRs) between 1972 and 

1983. The environmental gamma dose values (in the unit of mR) are presented for 11 facilities 

in environmental TBD Table 4-13: 

Monitoring locations, established for the 1950s and 1960s, are shown on 

Figure 4. As shown on this map, the 118 area film badge locations completely 

surround the IET testing area and effectively monitor gamma/beta releases from 

other facilities at the INEEL. Results of film badge monitoring for the 4th quarter 

of 1960 from 340 badges, pulled on a monthly basis, showed the average gamma 

reading to be <20 mrem/month and the beta to be <10 mrem/month. The 

maximum gamma was 40 mrem/month and the maximum beta was 

20 mrem/month, but locations for the maximum readings are unknown. Film 

numbers and environmental readings are provided in Table 7-1. (Peterson 

2004s, pg. 30) 

It should be mentioned that the above table gives data principally for off-site 

locations. However, especially during the early years, the data do include on-site 

locations. These locations were primarily along the highways; at the north end of 

the Site, along highways 22, 28, and 33, and at the south end of the Site, along 20, 

26, and along Lincoln Boulevard, which runs north-south through the Site. To 

illustrate, the H&S Division 1959 Annual Report (AEC 1960) briefly summarizes 

the onsite gamma monitoring program (pg. 120): (Peterson 2004s, pp. 30 and 32). 

Film badges were located at 30 stations throughout the NRTS as a means 

of area monitoring for external radiation. Fig. LIII shows the average 

monthly dose in millirem due to gamma radiation. Estimates of external 

doses downwind of ICPP as the result of the criticality incident of October 

16, 1959 were determined by this program. 

A 1970 environmental report (AEC 1970, pg. 3) states: (Peterson 2004s, pg. 33). 

Only the TLD’s located on the highways passing through the NRTS 

boundaries received exposures statistically above the background level. 

The highest exposure was 26 mrem for the six month period from 
 

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November 1969 through April 1970. Trucks carrying radioactive 

shipments on the NRTS highways are the likely cause of this exposure, 

which is 15% of the standard. 

For the period described, 69 TLDs at 69 locations were placed, retrieved, 

and read; 35 along highway 20 at the south end of the Site, 18 along 

highway 88 and 5 along highway 28 which are both at the north end of the 

Site, and 11 along highway 22 which runs north-south along the western 

lower half of the site. All TLDs, except those along highway 20, averaged 

20.9 mrem. Those along highway 20, the highway between CFA and 

Idaho Falls, Idaho, averaged 26 mrem. Since most radioactive shipments 

would come through Idaho Falls to CFA, the quoted statement about 

radioactive shipments is reasonable. 

In addition to the exposure readings described above, TLDs were placed around 

the major facilities at the NRTS and were used to measure the radiation exposure 

between 5/1 and 11/1/70. For example, 8 TLDs were placed around the EBORLPT facility with the highest reading of 19 mrem above background and the 

average being 12.9 mrem. Around the burial ground there were 18 TLDs with a 

highest reading of 760 mrem above background and an average of 257 mrem. At 

CFA, 4 TLDs averaged 16.8 mrem above background with the highest being 

20 mrem. At the TRA in 1970 there were 15 locations shown on Figure 6. At 

badge location #8 the badge reading was lost, but the other 14 readings averaged 

74.9 mrem. The highest reading was for location #10 (300 mrem), which is on the 

perimeter of the “north storage area” where contaminated or mildly activated 

items were stored. TLD locations #1 through #5 are along Monroe Blvd. where 

radioactive shipments come into and leave TRA and average 87.8 mrem. Of these 

5 TLDs, #3 and #4 are in one of two of the predominant wind directions and are 

most probably influenced by the effluent of the ETR and MTR stacks; they read 

110 and 130 mrem, respectively. Location #13 is practically adjacent to the ATR 

stack and reads 31 mrem above background. Badge #15, located on the fence 

southeast of the ATR reactor building, is at a location that would ‘see’ 

radioactive shipments coming into and leaving the TRA. Also, as radioactive 

shipment drivers are processing in and out of TRA, their shipments, sitting 

outside the guard gate, are influencing the exposure of badge #15. Also, the 

gamma irradiation facility is located just south of the TRA Guard Post #1, where 

the TRA badge rack is located. Badges #12 and #14 have the lowest readings, 16 

and 18 mrem, respectively, and are probably most representative of the higher 

exposures for personnel within the TRA not performing “radiation worker” work. 

Between the latter part of 1970 and the latter part of 1972, facility fence 

monitoring and facility locations had been established. Beginning in 1972 facility 

fence TLD measurements, made on a 6-month basis, are available in the 

Environmental Monitoring Data Reports (EMDRs) that were not made “public” 

but were sent to the contractor H&S managers. The data retrieved from these 

reports from 1970 to 1983 are tabulated in Table 7-2. For the years 1984 
 

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through 1992 facility fence TLD measurements could not be located, but for 1993 

and beyond, facility fence TLD measurements are again included in the 

INEL/INEEL EMRs. Background TLD measurements, corresponding with the 

facility fence TLD measuring periods, were also recorded in the respective EMR. 

All of the facility fence TLD data located is presented in Table 7-2. 

(Peterson 2004s, pg. 33) 

As shown in the table, all data for 1984 through 1992, is highlighted ... to 

indicate data that had to be added by extrapolation. For all facilities except TRA 

and ICPP, extrapolated data is the average of TLD data for that facility from 

1993 through 1999, “0s” excluded. For example, the extrapolated values for 

TAN-TSF is the average of four TLD-Bkgd values, 18, 4, 14, and 3, i.e., 39/4=10. 

For TRA and ICPP there is a downward trend in the readings with time. 

Therefore, for these two facilities, the bottom four values (1989 through 1992) are 

the averages of the 7 values below and the top five values (1984 through 1988) 

are the averages of the seven numbers directly above. Background values are 

provided in the last column of the table. (Peterson 2004s, pg. 34) 

First, NIOSH believes the fence-line TLDs adequately measured worker doses from direct 

beta/gamma radiation from the facility and also gaseous effluents released from the facility or 

from adjacent facilities (TBD, pg. 22). This is only valid, however, for the portion concerning 

“gaseous effluents released from adjacent facilities.” For “direct beta/gamma radiation from the 

facility” and for “gaseous effluents released from the facility,” the TLD measurements are not 

representative of what a worker would receive while working at or around the facility within the 

fence-line boundary. In fact, the TLD measurements would be too low to be representative since 

the fence-line TLDs were situated too far away from the locations of the workers and the source 

terms. Second, the TBD indicates that there was general contamination of the surrounding 

facility ground due to facility operations and release depositions. However, the TBD does not 

provide any data or instruction for dose reconstructions to account for this missed dose that 

would not be picked up by the fence-line TLDs. Third, NIOSH believes that it is unnecessary 

for the dose reconstructors to calculate missed dose due to facility air immersion from noble gas 

and halogen releases because they would have been measured by the fence-line TLDs. This 

assumption is not supported by any data. In addition, there are not enough fence-line TLDs to 

cover the facility ground to measure all potential air immersion releases. 

5.1.1.3.1 Validity of the Fence-Line TLD Direct Gamma Values 

Table 6-1 of the Occupational External Dose TBD presents the fence-line TLD direct gamma 

values for eleven facilities from 1952 to 2002. The values are calculated as mean values of 

several TLD measurements. Each of the TLD measurements has an uncertainty of one standard 

deviation. NIOSH should evaluate the resultant uncertainty associated with these values. 

Most important, a more recent INL Environmental Monitoring Fact Sheet, Direct Radiation 2004, 

shows different direct gamma values. In particular, for TRA, it presents much higher values 

from 2000 to 2002: 126 mrem for 2000, 172 mrem for 2001, 133 mrem for 2002, as compared 
 

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to 40, 27, and 34, respectively, from Table 6-1. These are significant differences. NIOSH should 

re-examine the direct gamma values, as well as the background values used in the TBD. 

5.1.1.3.2 Weakness of Fence-Line Monitoring Systems 

INL uses fence-line environmental TLD systems to monitor environmental releases from routine 

operational releases and episodic releases. These TLD systems are commonly used in all other 

DOE facilities and commercial plants, and, in general, are accurate and easy to use. However, 

they are intended for providing data for evaluating potential exposures to offsite populations 

around the INL area; they are not located or intended for providing exposure data for onsite 

workers. Therefore, the issue is whether these fence-line TLDs are measuring the environmental 

doses to personnel working at or around the facility where the release is originating. 

According to Table A1 of RAC 2002, the stack heights of various facilities at INL range from 

9.1 m (ARA) to 76.2 m (ATR and ICPP, respectively). With lower stacks, the possibility of 

onsite plumes during routine or episodic releases is higher. The worker exposures would be 

increased depending on the type and quantity of releases. NIOSH should evaluate this issue to 

determine the worker exposure to onsite plumes, especially since the fence-line TLDs would not 

be capable of identifying the doses to personnel who were working in the path of these groundlevel plumes within or around the facility. Mostly likely, the radioactive particulates of these 

ground level plumes would settle partially on the ground of the originating facility. Therefore, it 

is also necessary to evaluate the worker exposures to the resuspension of these ground 

depositions of the ground-level plumes and their accumulation over the operating years of the 

facility. For example: 

(1) For a ground-level radioactive halogen plume emitted at the facility where a worker was 

located, the fence-line TLD data would not be representative of the worker’s immersion 

doses. 

(2) As a facility ages, radioactive sources tend to accumulate around it, causing the general 

facility background to increase with time. The fence-line TLD data would not be 

representative of the actual direct gamma doses to workers working within the fence-line 

boundary because the workers were closer to the sources than the TLDs. 

(3) During the site interviews, INL management personnel indicated that they have soil 

sampling data, site survey reports, and facility baseline studies. NIOSH could develop a 

facility background dose rate within the fence-line boundary for each facility. 

(4) The INL environmental monitoring and surveillance program did not adequately monitor, 

evaluate, and report airborne effluents and ambient environmental conditions at INL 

(DOE-HQ 1991). The DOE-HQ Tiger Team observed deficiencies in the field sample 

controls, calibration and use of environmental monitoring equipment, and the use and 

storage of calibration standards. These deficiencies jeopardize the long-term reliability 

and integrity of the measurement process, and the validity of the monitoring data. 
 

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NIOSH should evaluate the representativeness of the fence-line TLD data for the onsite workers 

by comparing them to doses calculated from facility-specific on-ground radiation survey data, 

soil sampling data, and air sampling data. 

5.1.1.4 Other Observations 

5.1.1.4.1 Soil Sampling Data 

NIOSH did not evaluate soil-sampling data at different facilities to determine potential worker 

intakes from resuspension of radioactive materials deposited on facility grounds and fugitive 

emissions from radiologically contaminated soil piles. 

(1)	 As INL facilities age, cumulative depositions from effluent releases (both operational and 

episodic) on the grounds of each facility increase. Personnel working outside the facility 

within the fence-line boundary would have intakes from inhaling resuspended radioactive 

particulates. During the site interviews, INL management indicated that there are soil 

sampling data, aerial survey reports, site survey reports, and CERCLA reports that would 

provide such information. 

(2) Radiologically contaminated soil piles were observed at the ICPP, PBF, and ARA 

facilities by the DOE-HQ inspection team. These piles were enclosed by a rope barrier to 

prevent personnel from entering the area, but no mechanism was used to prevent 

windblown soil, as cited in a 1991 DOE Tiger Team Report (Vol. 1, pg. 3-69). 

(3) Dry evaporation, warm waste, and percolation ponds with radiologically contaminated 

soil could release fugitive emission, as cited in a 1991 DOE Tiger Team Report: 

The evaporation pond at the PBF was found to be dry during the 

inspection. It was noted that low-level radiologically-contaminated 

cooling tower water is released to the pond approximately twice a year. 

The pond would be dry for much of the year which may allow for wind 

blown contaminated soil being emitted from the pond. Other locations of 

dry ponds containing contaminated soil include the three warm ponds at 

the TRA, the percolation pond at ICPP, and the ponds at ARA. . 

(Vol. 1, pg. 3-70) 

According to Table A1 of RAC 2002, there are other ponds that may be major release 

points to the environment, including the following: 

a. ANL-W – 3 seepage ponds, leaching pits, and sanitary lagoon 

b. EBR-II – 1 industrial pond 

c. ARA – 2 surface depressions (1/3 acre and 1⁄2 acre) 

d. CFA – sewage plant tile drain field 

e. ICPP – 1 percolation pond and sewage plant to tile field 

f. LOFT – 1 pond 

g. NRF – 1 pond, 2 sewage ponds, and waste ditch 

h. PBF – 1 evaporation pond 
 

 

 

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i. TAN – 1 pond 

j. TRA – 2 ponds, 1 chemical waste pond and 1 leaching pond 

(4) Resuspension of radiologically contaminated soils in identified surface soil 

contamination at ICPP and other facilities due to wind blowing, personnel tracking, or 

vehicular activities, may cause significant worker intakes (Vol. 1, pg. 268). 

(5) Adequate protection for workers and the environment against the spread of radioactive 

contamination was not provided for known and potentially radiologically contaminated 

materials (or equipment) stored outdoors at INL facilities. Outdoor radioactive waste 

storage areas existed at TRA, PBF, RWMC, WERF, and CFA (Vol. 1, pp. 3-269, 3-271 

and 3-273). 

(6) The major contributor to dose at INL is radioactive gaseous releases. The dose fraction 

resulting from radioactive particulate releases during routine operations may not be 

accurate. Therefore, the calculated dose to workers may not be as accurate as could be 

achieved. Dose calculated as a result of unplanned releases may not be accurate also. 

This is cited as a finding in the 1991 DOE Tiger Team Report (Vol. 1, pg. 236). 

(7) Besides the contaminated soil piles, dry contaminated ponds, and contaminated 

equipment stored outdoors, several types of spills occurred at INL facilities that could be 

sources of resuspended radioactive material for worker inhalation exposure. During the 

interviews, site experts indicated that they had been personally involved in many spills of 

radioactive materials and liquids during transportation, transfer of materials, and 

maintenance works. 

5.1.1.4.2 Multiplying Factors 

The suggested multiplying factors for intakes given in the Occupational Environmental Dose 

TBD may not be claimant favorable for the following reasons: 

(1)	 The multiplying factor of two for intake values due to operational releases is based on a 

suggestion in the INELHDE Report. The authors of the INELHDE, however, further 

suggested that the uncertainty could be higher when the annual normalized ground-level 

concentration values are applied to the operational releases.4 

(2)	 The multiplying factor of 3 for intake values due to episodic releases was based on 

suggestion of the authors of the INELHDE Report. SC&A’s review of the INELHDE in 

2003 found the airborne releases associated with several of the Initial Engine Tests (IETs 

3, 4, and 10) of the Aircraft Nuclear Propulsion (ANP) Program were likely to have been 

underestimated as follows; (1) IET 3 – underestimate of total radionuclide release by up 

to a factor of about 3, (2) IET 4 – underestimate of noble gases by up to a factor of about 

4

 It should be noted that two of the authors of the INELHDE Report were also authors of the NIOSH INL 

Site Profile Report. 
 

 

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16, halogens by up to a factor of about 7, and solids by a factor of up to about 2, and (3) 

IET 10 – underestimate of total radionuclide releases by up to a factor of about 7. 

. 

(3) NIOSH performed an analysis to reduce the number of radionuclides from 56 to 9 for the 

operational releases. The 9 radionuclides retained contribute about 95% of the inhalation 

doses to the workers. However, it is not explained why this reduction is necessary, since 

a spreadsheet can readily calculate the inhalation doses for 56 radionuclides almost as 

fast as for 9 radionuclides. There seems to be no attempt in the TBDs to make the final 

inhalation doses claimant favorable by adding a multiplying factor of 1.05 to compensate 

for the missing 5% doses. While this 5% may not be significant in most cases, there are 

many small percentage take-aways in the TBDs that could add up to a significant amount 

when the claimant dose is approaching the 50% POC threshold. 

(4) The uncertainty of measurements made with film badges and TLDs could be as high as 

100%. It is not clear whether the dose reconstructor should use a multiplying factor of 2 

for direct gamma doses. In addition, NIOSH should address angular dependence of the 

gamma doses due to sensitivity of dosimeters used, as the procedures provided in OCASIG-001 Appendix B were found to be in error (SC&A 2005, pp. 143–145). It is essential 

for NIOSH to provide correction factors for dealing with angular dependence in the dose 

conversion factors used. 

(5) It is indicated in the Environmental TBD that the 1952–1962 values for direct gamma 

doses at TAN-TSF, TAN-LOFT, and TAN-LPT could be low by a factor of 3 (TBD, 

pg. 25). It is not clear whether the dose reconstructor should use a multiplying factor of 3. 

(6) For direct gamma doses after 1967, the uncertainty could be as high as 20%. It is not 

clear whether the dose reconstructor should use a multiplying factor of 1.2. 

5.1.1.4.3 Breathing Rate 

The NIOSH TBD (Section 4.2.1, pg. 10) uses an annual breathing rate of 2.4 x 103

 m3/yr rather 

than the 2.88 x 103

 m3/yr (1.6 m3/hr for 1800 hr/yr) of ICRP 68 or the 8.0 x 103 m3/yr (RAC2002, 

pg. 38) of the NCRP. The TBD assumption appears less claimant favorable than the ICRP or 

NCRP assumptions. However, NIOSH believes the chosen breathing rare is applicable to a 

worker exposed to the environment, who is not breathing hard from strenuous labor. NIOSH 

also believes that this choice of breathing rate was made uniformly for consistency for all the 

evaluated sites. This consistency choice is not claimant favorable for INL claimants. There is 

also no data supporting the claim that INL workers exposed to outside facility environments are 

not breathing hard from strenuous labor.5 

5

 At the time of the preparation of this report, SC&A was informed that NIOSH has addressed the issue as 

part of the latest revision of the Bethlehem Site Profile Report. 
 

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5.1.1.4.4 Naval Reactor Facility Data 

Although exposures to workers at the Naval Reactor Facility (NRF) located on the INL site are 

outside the scope of the dose reconstruction project, operations at the NRF have the potential of 

exposing personnel located outside the facility. Release data from the Naval Reactor Facility 

was deemed classified until recently. The INELHDE Report assumed values for environmental 

release from this facility (Attachment 3) and identified four NRF releases (1955, 1975, 1976, 

1977) (DOE 1991a, pp. A-52 and A-55). In 2004, NRF provided environmental release and 

emission information to the INL site. NIOSH did request NRF environmental release documents 

from NRF more than 1 year ago. These documents were redacted and ready to be picked up by 

Intrepid, a subcontractor to ORAU, the NIOSH dose reconstruction contractor. NIOSH has since 

indicated no further interest in this data. SC&A believes that NIOSH should review this newly 

released data to improve estimation of the intake values and improve understanding of the origin 

of personnel exposure from NRF. 

5.1.1.4.5 Background Dose Subtraction 

It is confusing to compare INL fence-line direct gamma values provided in Table 4-13 of the 

environmental TBD with values given in Table 7-2 of the supplement to the TBD (Peterson 

2004s). The background values in Table 4-13, which are supposed to be subtracted out from the 

TLD values, are higher from 1970 to 1983; NIOSH does not explain of the differences. To 

achieve claimant-favorable conditions, NIOSH should use the lower background values. In 

addition, a more recent INL Environmental Monitoring Fact Sheet, Direct Radiation 2004, 

shows some different direct gamma background values for 2001, 2002, and 2003. 

5.1.1.4.6 Releases of Uranium and Its Daughters 

The Occupational Environmental Dose TBD indicates that an analysis was performed to reduce 

the number of radionuclides from 56 to 9 for the operational releases. Actually, the TBD 

reduced the number of radionuclides from a total of 71 to 9 listed in the INELHDE for the 

operational releases. It also indicates that the dose contributions from these 9 radionuclides 

would represent 95% of the inhalation dose. Similarly, the TBD reduces the number of 

radionuclides from a total of 52 to 15 for criticality events, 9 for special tests, and 7 for IETs. 

However, NIOSH does not provide any compensations or adjustments to the lost 5% 

corresponding to operational releases or whatever amount corresponding to the episodic releases. 

As a result of these reductions in the radionuclides considered, uranium and its decay daughters 

are not included in the worker intake value tables. The concern may be less significant if we 

consider only the dose impact to workers at a facility distant from the (release) source facility. 

However, as already pointed out in the above discussions, the workers would receive 

considerably higher doses from inhaling released radioactive material in a ground-level moving 

plume or resuspended contaminated material from ground deposition at the facility where the 

workers were located. Inhalation, ingestion, or oro-nasal intake of uranium and its daughters 

might then be a significant issue. NIOSH should re-examine the potential missed worker dose 

due to intake of uranium and its daughters. 
 

 

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5.1.2 Occupational Internal Dosimetry Issues 

The NIOSH INL occupational internal dosimetry TBD reflects an extensive and thorough review 

of selected existing documents related to INL internal dosimetry programs, bioassay records, and 

reports. NIOSH interviewed INL Dosimetry Department staff and requested individual 

dosimetry records and documentation. NIOSH also interviewed some workers and retirees as 

part of an outreach program after the TBD was done. However, NIOSH (apparently) did not 

look at or use facility-specific field logs, RWPs, PEQs, facility survey reports, incident reports, 

occurrence reports, contamination reports, and ALARA records. 

The authors of the Occupational Internal Dose TBD display confidence in the past and current 

INL radiological protection programs and its implementation, the accuracy of the internal 

dosimetry programs including whole-body counting and bioassay program (both routine and 

event specific), and the dosimetry record-keeping systems. During the site interviews, the INL 

Dosimetry staff showed similar confidence: 

Except for several criticalities, there has been no significant radiological control 

incident for more than 55 years of operations. The site records anomalies, 

personnel contamination, skin exposure, and over-exposure incidents on a 

Personnel Exposure Questionnaire (PEQ) form. For example, a lost dosimeter by 

a worker would trigger a PEQ process. Anyone onsite can issue a PEQ, but they 

are usually generated from the Radiological Control Organization. In the case of 

skin exposure evaluations, the PEQ will be placed in the dosimetry record. Other 

PEQs are stored together and are not placed in the dosimetry file. If there was a 

dose adjustment as a result of a PEQ evaluation, this would be noted in the 

electronic dosimetry database. The PEQ itself is not provided to NIOSH as a part 

of the claimant packet. 

Personnel contamination reports are maintained as a field record. Only the 

exposure assessment is maintained in the dosimetry record. This includes 

primarily dosimetry information such as organ doses, radioisotopes of concern, 

total doses for the events, and bioassay results. 

It is important to point out here that many records and data pertinent to the worker claims are not 

included historically in the workers’ files. These potentially missing records and data include 

personnel contamination reports, PEQs, RWPs, incident reports, and occurrence reports. For 

example, in the DOE record of some workers, the only indication of an incident is a brief 

comment on the bioassay report. No PEQ, incident report, or personnel contamination record are 

available to ascertain the details of the incident that may affect the dose reconstructions; this is 

apparent, for example, in two case reports that SC&A examined.6,7 

Based on historical internal exposure source terms and known exposure incidents (both in-vivo 

and in-vitro records) at 13 specific INL facilities, the TBD identifies radionuclides of concern for 

6

 Case No. 11654 from Idaho National Laboratory, Dose Reconstruction in progress. 

7

 Case No. 10689 from Idaho National Laboratory, Dose Reconstruction Approval Date February 26, 2004. 
 

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key INL facilities and develops potential inhalation intakes for workers. Among these inhalation 

intakes, NIOSH recommends a set of defaults for specific radionuclides (such as Cs-137, Sr-90, 

Ce-144, Pu-238, or Pu-239) with weighting factors for the missed dose calculations based on 

urine, gross beta, in-vivo, and other bioassay data. The dose weighting factors are intended to 

account for missed inhalation dose from the other radionuclides. However, these weighting 

factors may not be claimant favorable, because there are no uncertainty evaluations for the 

missed doses and their weighting factors. 

NIOSH’s missed dose defaults are developed for four time periods; (1) start up through 1960, 

(2) 1961–1970, (3) 1971–1980, and (4) 1981 to present. For the first two time periods, NIOSH 

recommends defaults for two components: (1) workers at all INL facilities; and (2) workers at 

TRA. For the third time period, NIOSH recommends defaults for five components; (1) workers 

at all INL facilities, (2) workers at INTEC and unknown locations, (3) workers at ANL-W, 

(4) workers at other locations, and (5) workers at TRA. For the last time period, NIOSH 

recommends defaults for four components; (1) workers at all INL facilities except ANL-W, 

INTEC, and SMC, (2) workers at ANL-W, (3) workers at INTEC and unknown locations, and 

(4) workers at SMC. These missed dose defaults are inhalation doses only. There were no 

considerations given for potential ingestion or oro-nasal doses for uptakes due to unplanned 

events or specific exposure incidents. 

The missed internal dose defaults are presented in Table 5.7-1, and are to be applied to dose 

reconstructions according to the following condition: 

If the bioassay records do not include the radionuclide analyses and only record 

gross beta or alpha results, default assumptions are described in the following 

text (pg. 8). 

NIOSH’s key assumption for dose reconstruction of missed dose is as follows: 

If claimant file includes positive external dosimeter readings, they should be 

treated as radiation workers and the default internal missed dose is applied as 

outlined in the table. If no detectable external or internal dose information is 

recorded, only the environmental dose should be included. (TBD, pg. 37) 

Another key assumption that NIOSH makes in the TBD is that claimant files are all complete 

with all documented incidents and internal doses included. NIOSH believes that the inclusion of 

these outlined missed doses would fill any gap in the information for undocumented missed 

worker internal doses. 

In addition, the TBD specifies a guideline for dose reconstruction: 

When there is no evidence in the incident file or the individual’s dosimetry file 

that an individual was involved, and no other supporting evidence supporting that 

an individual was involved in the accident, it should be assumed that the 

individual was not involved in the incident. (TBD, pg. 8) 
 

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It is ironic that the above guideline dictates how the completeness or the incompleteness of the 

incident file or the individual’s dosimetry file would become conclusive evidence supporting that 

an individual was not involved in the accident. But, at the end of the TBD, NIOSH amends its 

guideline somewhat to read the following: 

Thus the probability that a worker received a significant unmonitored internal 

intake of radioactive material is very low. It is recommended that workers who 

have no recorded internal dose and wore a personal dosimeter be treated the 

same as a worker who was monitored but had no bioassay results exceeding 

reporting levels. It is further recommended that individuals that were not issued a 

personal dosimeter and have no record of internal dose monitoring be assigned 

only the environmental dose for the facility. (TBD, pg. 39) 

The following enumerates and comments on some of the pertinent information NIOSH has 

included in the internal dose TBD: 

(1) NIOSH evaluates recorded internal doses for 1992–2000 in Table 5.1.4-1. This table also 

provides radionuclides of concern at the major INL facilities in recent years, but it is 

quite inadequate for dose reconstructions, with almost 42 years of missing internal dose 

data from 1949 to 1991. It also does not provide radionuclides of concern for all INL 

facilities. Furthermore, there are no instructions for dose reconstructors on how to use 

these internal dose data. 

(2) NIOSH summarizes the primary radionuclides of concern at INL in Table 5.1.4-2. This 

table also lists the default solubility class and identifies the preferred bioassay techniques 

for the radionuclides. The TBD indicates that the default assumption of M or S should be 

used, based on the most claimant-favorable result to the organ in question. 

(3) NIOSH summarizes the history of internal dosimetry efforts at INL and concludes the 

following: 

The largest internal exposures at INEEL have resulted from accidental 

intakes associated with episodic events or planned major releases, for 

which the time and characterization of the materials of the intakes were 

well known. These exposures were documented in each exposed 

employee’s file. (TBD, pg. 8) 

Most internal doses have been identified following an incident rather than 

as a result of routine bioassay measurements. (TBD, pg. 9) 

The internal dose reconstruction for personnel who have worked at a 

number of INEEL facilities should rely on specific bioassay data 

(radionuclides, quantities, etc.) when available. (TBD, pg. 12) 
 

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These conclusions lead to NIOSH’s missed internal dose approach. However, they are 

consistent with the findings of the DOE-HQ Tiger Team and DNFSB audits, as well as 

the information provided by the site experts interviewed. 

(4)	 NIOSH believes that there has been “a basic level of site wide consistency in the internal 

dosimetry programs applied to INEEL facilities and programs, and particularly the 

bioassay analytical techniques and calculational processes” (TBD, pg. 12). However, this 

conclusion is not consistent with the information provided by the site experts interviewed. 

(5)	 NIOSH believes that the internal dose records at INL are adequate and effective to 

support the dose reconstruction process. This is again not consistent with the findings of 

the DOE-HQ Tiger Team and DNFSB audits, as well as the information provided by the 

site experts interviewed. 

(6) NIOSH summarizes the routine bioassay history at the INL site in Table 5.2.2-1, which 

provides the type of bioassay (in-vitro, in-vivo, urine, or fecal), frequency, group of 

worker sampled, investigating level, and sources of information. It covers the operating 

years of INL from 1953 to 2001. The table shows that individual data sheets were 

reviewed to determine the bioassay programs between 1953 and 1981. After 1981, 

several technical papers were used as the sources for bioassay program information, and 

it is explained in the TBD why individual data sheets were not reviewed. It is also 

interesting to point out that the RCIMS database was not referenced or used in any way in 

this table. 

(7) NIOSH believes the INL internal dose information, including the calculated internal 

doses as well as the in-vitro and in-vivo individual bioassay results, is in full compliance 

with Federal regulations. NIOSH believes “all (negative as well as positive) bioassay 

data were recorded in the individual dosimetry files” (pg. 15). This presumption leads to 

NIOSH’s approach in this TBD, but is inconsistent with the findings of the DOE-HQ 

Tiger Team and DNFSB audits, as well as the information provided by the site experts 

interviewed. 

(8) The TBD indicates that, “during the early years internal dose was usually considered 

separately from external dose in terms of meeting specific exposure limits and the 

calculated dose was only reported and documented if specific dose level were exceeded 

(AEC/ERDA 0524, 1968–1977, required periodic urinalyses and/or in-vivo counting 

and/or evaluation of air concentrations if the whole-body dose or dose commitment could 

exceed 300 mrem in a calendar quarter). Changes in the reporting levels did not 

generally result in changes to the air monitoring and bioassay sampling programs. Each 

individual analytical result was documented and placed in individual exposure files 

regardless of the formal reporting requirements” (TBD, pp. 15–16). NIOSH believes and 

concludes here that the air monitoring programs, bioassay sampling programs, and 

worker exposure record-keeping programs at INL are consistently effective and complete. 

This conclusion again is inconsistent with the findings of the DOE-HQ Tiger Team and 

DNFSB audits, as well as the information provided by the site experts interviewed. 
 

 

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(9)	 NIOSH lists the possible internal dose assessment information that may be in claimant 

files after 1989 in Table 5.2.4-1, and before 1989 in Table 5.2.4-2. There are many 

differences in the details of the information provided between the two periods. The TBD 

also provides other analytical information in Table 5.2.4-3 and area coding used in the 

claimant dose files in Table 5.2.4-4. In Table 5.2.4-5, NIOSH lists the derived 

investigation levels (μCi) in 1977 for 15 primary radionuclides of concern at INL for 

acute exposures, which represent one-tenth of the then quarterly radiation standard. 

However, the TBD does not provide guidance on how to fill the data gap if any 

information essential to the dose reconstruction is missing from the claimant’s files. 

(10) NIOSH presents a history of surface contamination control and MDAs used at INL, 

including an effective air monitoring program with the use of CAMs, gamma counting of 

bioassay samples, urinalysis, fecal sample analysis, and whole-body counting. 

Table 5.3-1 of the TBD lists the control levels and detection techniques for surface 

contamination and corresponding MDAs for INL between 1952 to present. However, 

NIOSH states the following: 

The consistent NRTS/INEEL policy and practice was to require 

respiratory protection on jobs when the possibility of generating airborne 

contamination was thought to exist, regardless of the actual measured air 

or surface contamination. These practices influence the assumptions 

related to dose evaluation in internal dose reconstruction. (TBD, pg. 18) 

NIOSH should perform uncertainty analyses on the contamination control equipment and 

analytical capabilities at INL. The uncertainties due to the sensitivities of the equipment 

and practices are essential for dose reconstruction, especially for the early days of the lab 

when the equipment was not as accurate and the program practices were less rigorous. 

(11) NIOSH believes “workers were asked to submit to bioassay whenever they were in an 

area in which a CAM alarmed” (TBD, pg. 19). NIOSH should review site audit and 

inspection reports by DOE-HQ (Tiger Team), DOE-ID, and DNFSB. A 1991 Tiger 

Team report found the set point for CAM alarms was faulty. Site experts interviewed 

indicated that the practices at some INL facilities frequently ignored CAM alarms. 

(12) NIOSH presents some urinalysis results for 1959, 1960, and 1961 in Table 5.3.2-1 of the 

TBD, and summarizes the numbers of urinalyses performed and the highest results for 11 

primary radionuclides of concern, as well as gross beta and gamma exposures. These 

urinalyses were limited to data from 1959, 1960 and 1961. NIOSH should provide more 

information for post-1961 years that constitutes almost 45 years of site operations. 

(13) Instead of evaluating the analytical sensitivities of the INL bioassay equipment and 

adequacy of procedures, NIOSH applies environmental sample analysis detection limits 

and sensitivities for water (alpha, beta, and tritium) and milk (I-131, Sr-90) to INL 

special and routine bioassay sample analyses. The TBD lists the detection limits 

applicable to environmental sample analyses from 1953 to 1965 in Table 5.3.2-2. It is 

not clear how dose reconstructors should use these data; should they also apply them to 
 

 

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post-1965 analyses? NIOSH should provide equipment-specific and method-specific 

information. The broad application approach taken lacks technical validity, as well as 

data support. Not only is it limited to the period of 1953 to 1965, but also equipment and 

practices changed significantly after 1965. NIOSH should perform uncertainty analyses 

on the bioassay equipment and analytical capabilities at INL. The uncertainties due to 

the sensitivities of the equipment and analytical methods are essential for dose 

reconstructions, especially for the early days of the site when the equipment was not as 

accurate and the analytical methods were less rigorous than they became in later years. 

(14) In TBD Table 5.5-1, NIOSH presents a single set of whole-body counting summary 

statistics for INL workers in 1963, showing the number of times reported, number of 

individuals, and maximum activity (μCi) for 24 radionuclides of concern at INL. The 

TBD states that, “the maximum activity detected provides an upper bound on how large 

an activity might be found in someone in earlier years before the whole-body counter was 

operational” (TBD, pg. 23). Since the table only shows data from 1 year of operations, 

NIOSH should compile data for all post-1963 whole-body counting statistics. In addition, 

NIOSH has stated repeatedly in the TBD that the worker exposures were comparatively 

worse in the early years, due to the nature of the operations and the improvement in 

instrumentation, program standards, and methodologies. Therefore, the maximum 

activity detected in 1963 may not represent an upper bound for exposures occurring in 

earlier years. 

(15) NIOSH evaluated 13 specific facilities to identify potential sources of internal exposures 

based on radionuclide activities from facility operations. 

• Test Area North (TAN) 

• Specific Manufacturing Capability Project (SMC) 

• Idaho Chemical Processing Plant (ICPP, now INTEC) 

• High-Enriched Spent Fuel Storage (CPP-603) 

• High-Level Waste Storage Tank Farm 

• High-Level Waste Calcination 

• Process Analytical Facilities 

• Spent Fuel Processing 

• Argonne National Laboratories – West (ANL-W) 

• Radioactive Waste Management Complex (RWMC) 

• Waste Reduction Operations Complex (WROC) 

• Test Reactor Area (TRA) Reactors 

• Test Reactor Area Laboratories 

Similar to the Occupational External Dose TBD, NIOSH should provide information on 

airborne concentrations and contamination levels in these facilities. It would be even 

better if NIOSH could develop a list of high-risk (dose) jobs in each of these facilities. 

(16) NIOSH indicates that detailed particle size analyses were performed for SMC operations. 

The internal dose TBD states, “an AMAD of 2.4 μm is appropriate for typical SMC 

operations” (pg. 29), but NIOSH does not provide support information for this value in 
 

 

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the TBD. In addition, particle size analyses are not provided for the other 12 facilities. 

This omission could significantly impact the missed dose calculations for dose 

reconstructions. It is, otherwise, interesting to note that the Occupational Environmental 

Dose TBD indicates that, since there was no particle size analysis performed at the INL 

facilities, default values should be used for worker dose reconstructions. The site experts 

interviewed indicated that a particle size study was completed for the calciner with a 

cascade impactor at ICPP by the field organization. Particle sizes determined range from 

0.3 to 1 μm. There is no other particle size study recently. The default particle size for 

internal dose calculations is 1 μm AMAD. 

(17) NIOSH lists the depleted uranium (DU) mass and activity ratios for SMC in 

Table 5.6.1-1. The TBD also indicates that a baseline uranium background concentration 

study was performed to represent “nonoccupational elimination of the SMC worker 

population” (TBD, pg. 29). This study used urine samples from SMC non-radiation 

workers in 1987, 1994, and 1998 to determine the nonoccupational component of 

uranium excretion for SMC radiation workers. The value of 0.16 μg/L was calculated 

and subtracted from each worker urine sample result prior to assessment of the 

occupational internal dose. This approach is not statistically valid, due to the nature of 

the selected nonoccupational worker group, collocated with the radiation workers. There 

was an obvious possibility that these non-radiation workers may have intakes of uranium 

from the surrounding contaminated environment. NIOSH should evaluate the validity of 

the value of 0.16 μg/L used as nonoccupational subtraction from urine results. More 

importantly, the background concentration should not be subtracted from the urine 

sample results. (See Section 5.1.2.8 for further discussion). 

(18) NIOSH lists nine notable airborne incidents at ICPP in Table 5.6.2-1 of the TBD. The 

table provides the date of the incident, radionuclides released, internal dose results, and 

references. However, the TBD concludes that, “The primary internal dose experience at 

the ICPP resulted from accidental releases” (pg. 29). This conclusion seems to disagree 

with audit findings, and past and present worker experiences. NIOSH should re-evaluate 

the potentials of internal dose exposures at ICPP by examining high-risk (dose) jobs and 

field records. 

(19) NIOSH indicates that for personnel who worked for extended time periods in the 

Building 603 storage facility at ICPP if, “specific bioassay analyses are either not 

available or insufficient, a claimant favorable default intake of 1000 DAC-hours per year 

should be assumed” (TBD, pg. 30). There is a tendency for the TBD authors to use 

average values for determining default values. In almost all (valid) cases of the dose 

reconstruction, the missed doses are in the high end of the dose spectrum. Therefore, 

NIOSH should provide both average and upper bound values for defaults to assist the 

dose reconstructors. For example, NIOSH could use the 95th percentile values for those 

workers in high-risk (dose) jobs. 

In this case, the 1,000 DAC-hour is assumed in the TBD to be the average intake by a 

worker in Building 603, with Cs-137 and Sr-90/Y as the primary radionuclides of 

concern. This assumption is based on a 1 MPC level and 1,000 hours per year, according 
 

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to NIOSH experts (see Attachment 1). This assumption would short-change some 

workers who may have worked in the building for more than 1,000 hours or in a very 

highly contaminated area, receiving the 1,000 DAC-hour in a short period of time. In 

addition, the default airborne activity of 2 x 10-9 μCi/cc is not justified by supporting data 

and uncertainty analyses. For example, a pipefitter who worked in Building 603 on a 

valve or piping replacement job could be working in a contaminated air zone with much 

higher concentration levels than this default airborne activity value. Therefore, this 

assumption does not appear to be claimant favorable. 

(20) There were no potential internal exposure source term data identified for the High-Level 

Wastes area, the High-Level Waste Calcination area, and the Process Analytical Facilities 

at the ICPP. As the TBD and site experts interviewed indicate that the waste calciners 

(both WCF and NWCF) were among the worst contaminated facilities at INL, NIOSH 

should have identified the high-risk (dose) jobs and potential missed doses to personnel 

who worked in these facilities. 

(21) NIOSH identifies three types of spent nuclear fuels (aluminum-clad, stainless steel-clad, 

and zirconium-clad) as the primary source terms for worker internal exposures at ICPP. 

The TBD further determines that the radionuclides of concern are the “most limiting 

radionuclides” for all the facilities at the INL site. (Eventually, the TBD uses these 

radionuclides as the default limiting radionuclides to calculate missed internal doses for 

INL workers in Table 5.7-1). The radiologically significant radionuclides associated with 

these fuels are listed in Table 5.6.2.5-1. The inhalation doses from inhaling radioactivity 

from these three types of fuel were calculated using ICRP 68 5 μm AMAD dose 

conversion factors. NIOSH selected four key radionuclides as the main dose contributors 

for potential internal exposures; Sr-90, Cs-137, Ce-144, and Pu-238/Pu-239 

(interchanged among fuels). Three weighting factors were developed by ratios and used 

to account for the dose contributions from unused radionuclides: 

Table 5.6.2.5-1 contains too many radionuclides for efficient dose 

reconstruction. Rather than include all of the radionuclides in the default 

summary table for missed dose (Table 5.7-1), only Sr-90, Cs-137, Ce-144, 

and Pu-238 are included for aluminum and zirconium fuels. For stainless 

fuels, the Pu-238 is replaced by Pu-239. Cesium-137 was selected 

because it is most commonly reported in in vivo results rather than for its 

dose contribution. The potential missed inhalation dose from the other 

radionuclides is accounted for by weighting the dose from these selected 

radionuclides by the weighting factors at the bottom of Table 5.6.2.5-1. 

This gives an equivalent to 100% of the dose from the radionuclides 

distribution for the three types of fuels. (TBD, pg. 33) 

Again, NIOSH’s approach used here is based on the “average” percent inhalation doses a 

worker would receive due to the “average” relative activities of the identified significant 

radionuclides of concern for the ICPP processed fuels. It does not take into account of 

the upper bound inhalation doses. 
 

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(22) NIOSH indicates that there is one exception to the above spent fuel approach; the RaLa 

(Radioactive Lanthanum) process, in which large quantities of volatile radioactive 

iodines (i.e., I-131, I-132, and I-133) were released. However, the TBD does not provide 

instructions to dose reconstructors on how to evaluate missed doses for personnel 

working in the RaLa process or how to scale from the spent fuel data. 

(23) NIOSH identified and evaluated radionuclides of concern at ANL-W, RWMC, and 

WROC. Analytically determined MDAs, bioassay data, and radioactivity concentrations 

are also provided in the TBD (Tables 5.6.3-1, 5.6.4-1, and 5.6.4-2). The TBD also states 

that, “a comprehensive radiation protection program is practiced, which includes 

extensive air monitoring, personnel contamination, and surface contamination 

surveillance” (TBD, pg. 34). Once again, the TBD shows a degree of confidence in the 

INL radiation protection programs that is not consistent with the findings of the DOE-HQ 

Tiger Team and DNFSB audits, as well as the information provided by the site experts 

interviewed. 

(24) NIOSH concludes that, “For most of the history of the INEEL personnel dosimeters were 

issued to all workers at facilities handling radioactive material” (TBD, pg. 37). The TBD 

continues to show the sentiment that the authors have full confidence in the radiological 

protection policy and programs and internal exposure monitoring systems at INL. 

NIOSH believes, “the probability that a worker received a significant unmonitored 

internal intake of radioactive material is very low” (TBD, pg. 37). 

5.1.2.1 Completeness and Quality of INL Internal Dosimetry Programs 

The authors of the Occupational Internal Dose TBD display full confidence with the radiological 

protection programs, the internal dosimetry programs, and the dosimetry record-keeping systems 

at INL. For instance, NIOSH states the following in the TBD: 

As a consequence of a consistent AEC/DOE policy to avoid detectable internal 

exposures, coupled with the time and technical complexity of an internal dose 

evaluation, the general policy at INEEL for internal exposure has been preventive 

in nature. (TBD, pg. 18) 

The consistent NRTS/INEEL policy and practice was to require respiratory 

protection on jobs when the possibility of generating airborne contamination was 

thought to exist, regardless of the actual measured air or surface contamination. 

These practices influence the assumptions related to dose evaluation in internal 

dose reconstruction. (TBD, pg. 18) 

NIOSH also believes that all worker exposures were documented in each exposed workers’ file. 

As a result, a dose reconstructor could perform the claimant dose reconstruction based on 

information kept in the claimant file. These conclusions are evident throughout the TBD and can 

be illustrated by a few excerpts: 
 

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Most internal doses have been identified following an incident rather than as a 

result of routine bioassay measurements. (TBD, pg. 9) 

The largest internal exposures at INEEL have resulted from accidental intakes 

associated with episodic events or planned major releases, for which the time and 

characterization of the materials of the intakes were well known. These exposures 

were documented in each exposed employee’s file. (TBD, pg. 8) 

Each individual analytical result was documented and placed in individual 

exposure files regardless of the formal reporting requirements. (TBD, pg. 16) 

However, information provided by past and present workers during SC&A’s site interviews, as 

well as INL site inspection findings by the DOE-HQ Tiger Team in 1991 and by the Defense 

Nuclear Facilities Safety Board in 1994 and 1995, indicates that the INL site programs were 

deficient in many ways, despite what was claimed. These program deficiencies are serious and 

likely jeopardize the validity of data NIOSH used and its conclusions in the TBD. There are 

many examples of potentially important deficiencies, which could adversely affect the dose 

reconstruction process, cited by the DOE Tiger Team, as follows (DOE-HQ 1991): 

(1) The Radiological, Environmental and Safety Laboratory (RESL) was in charge of all air 

sampling and bioassay analysis and internal dose assessment functions at INL, but RESL 

did not have an effective safety and health management system. The Tiger Team found 

75 safety deficiencies in policy, policy implementation, and procedures. These 

deficiencies put all the analytical results RESL produced in question. For example 

(Vol. 2, pg. 4-26): 

a. The calibration of whole-body counter and teletherapy sources were not adequate 

(Vol. 2, pg. 4-467). 

b. RESL was found to be deficient in monitoring airborne radioactivity in laboratory 

spaces, in performing pre-employment baseline bioassay for new employees, in 

performing bioassay surveillance of individuals potentially exposed to airborne 

radioactivity, in performing fume hood testing, in performing quality assurance 

sample checks for routine urinalysis, and in performing analysis of urine samples 

for total uranium from INL radiation workers (Vol.2, pp. 4-613 and 4-614). 

c. RESL was deficient in ensuring accurate documentation of the individual radiation 

exposure, including routine survey information (instrument type and number and 

reading), wipe sample, counting, and air sample counting (Vol.2, pp. 4-623 and 

4-624). 

(2) The INL Internal Dosimetry Program was found to be deficient because compliance with 

DOE 5480.11 (radiation protection) could not be demonstrated (Vol. 2, pg. 4-181). 

a. For sample implementation of in-vivo and in-vitro bioassay programs, logs were 

not maintained to provide information such as purpose of bioassay, schedule of 
 

 

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bioassay, date when bioassay was completed, data when the analytical results were 

received, analytical results, and date a dose assessment was completed. 

b. For workers, required in-vivo and in-vitro bioassay monitoring was not completed 

as scheduled. This was primarily due to the cumbersome process of dose 

assessments involving four different entities, namely RESL, INL Laboratories, 

RSRS, and ODU. In fact, RESL was not in compliance with DOE Orders. 

(3) INL was deficient in providing adequate calibration equipment or facilities to perform 

tests for insuring reliability of existing radiation protection instrument response in 1991 

(Vol. 2, pg. 4-183). 

a. There were no defined criteria for acceptance testing of fixed radiation protection 

instruments. 

b. Trending analysis of radiation protection instrument performance was not 

performed. 

c. There was no program to identify problems and improve the response of radiation 

protection instrumentation to meet the needs of specific applications. 

(4) Several deficiencies of the CAMs were identified at TAN, ATR, and other facilities. 

Their locations were not based on an evaluation of airflow patterns (Vol.2, pg. 4-185). 

a. Some of the CAMs detecting beta-gamma radiation had a set point of 

approximately 8 DAC-hour instead of 1 DAC-hour (i.e., MDL). 8 DAC-hour was 

the minimum set point for alpha emitters but not for beta-gamma. 

b. Some CAMs (e.g., TAN Warm Shop) were not in service. There were no two level 

alarm systems established for the CAMs. 

(5) Maintenance of the INL occupational exposure record was transferred from the RESL in 

1988. The quality of the historical occupational exposure records ranges from 

handwritten in pencil on plain paper to records legibly recorded on standard forms in ink 

or using computer-generated data. The records of individual occupational exposure 

histories are readily retrievable. The histories can be obtained from ODU (i.e., manual 

retrieval from record storage sites). Information prior to 1986 must be retrieved from a 

combination of microfiche, notebooks, and individual files. Information from 1986 

forward is readily retrievable from computer record systems. Documented procedures for 

retrieval of occupational exposure histories are maintained. The results of in-vivo and invitro bioassay are maintained. However, there are certain limitations in the exposure 

records and the process of incorporating records in the dose reconstruction files. These 

limitations may hinder dose reconstructions and result in missed doses. Overall, the 

Tiger Team found the maintenance and retention of occupational exposure record was 

deficient at INL. For example (Vol. 2, pp. 4-190 to 4-192): 
 

a. The results of bioassay are not included in an individual’s file unless a dose 

assessment was made from a positive result. 

b. The individual’s file does not indicate that negative bioassay results are available. 

c. Records of dose estimates from special studies and unusual exposures are not 

maintained in an individuals’ occupational exposure history. 

d. There was no audit mechanism to verify that previous occupational exposure 

histories, exposures of employees who travel to other sites and internal dose 

assessments are included in the files. 

e. Reporting of occupational exposure to radiation workers was deficient in the areas 

of submitting annual report of radiation exposures, using the effective dose 

equivalents, and providing summary of annual, cumulative and committed effective 

dose equivalents to each radiation worker on an annual basis. 

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(6) INL was deficient in routine and accident personnel radiation dosimetry program at SMC 

(Vol. 2, pg. 4-611). 

(7) The procedure for personnel entry into potentially airborne radioactivity areas when 

wearing respirators was deficient at INL because no air sample measurements were taken 

to assure that respirator protection factors are adequate (Vol. 2, pg. 4-612). 

(8) The INL site was found to be deficient in testing and calibration of radiation 

measurement instruments including gas proportional counters, air sample counters, 

portable radiation survey and frisking instruments, radiation counters for wipes, special 

tool survey box instruments, and portal monitors (Vol.2, pp. 4-615 and 4-616). 

(9) The INL radiological air sampling program was deficient in ensuring timely and 

representative airborne activity measurements in the workplace (Vol.2, pp. 4-617 and 

4-618). 

In addition to the many critical findings by the DOE-HQ Tiger Team, the DNFSB also generated 

a long list of deficiency findings on the INL radiological protection programs and internal 

dosimetry programs during their audits. Several examples of the DNFSB findings, which may 

impact dose reconstruction, follow: 

(1) 7/29/1994 DNFSB finding: The staff noted an inconsistency among contractors in the 

requirements for respiratory protection, as well as some errors in procedures. 

(2) 7/29/1994 DNFSB finding: An error was noted in EG&G Procedure 10.3, Airborne 

Radioactivity Monitoring, dated January 28, 1994, and Document Revision Request 

(DRR) dated February 14, 1994. These documents gave equations for computing High 

Alarm set points that are not correct. 
 

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(3) 7/29/1994 DNFSB finding: Air monitoring for alpha radioactivity in the building is 

accomplished by one continuous air monitor. 

(4) 7/29/1994 DNFSB finding: Although several individuals are involved in operations in 

the high-level waste pit area where dose rates can reach several rem per hour, standard 

actions are not taken to control personnel exposure and preclude unnecessary exposure. 

(5) 9/15/94 DNFSB finding: ICPP radiological control practices also remain deficient, 

including a general lack of concern with personnel and equipment crossing radiological 

control boundaries, inadequate frisk/swipe survey practices, etc. 

(6) 1/17/95 DNFSB finding: Radiological control postings and field documentation on the 

tank farm were inconsistent and contained mistakes that were not identified and/or 

corrected by internal reviews. 

(7) 1/17/95 DNFSB finding: Several protective clothing doffing areas do not have posted 

doffing instructions. In one location, a Radiation Area posting was changed by hand to 

read "High Radiation Area." No radiation levels were listed. Radiation levels were 

entered inconsistently on many postings throughout the project. 

Given these deficiencies noted in the INL radiological protection and internal dosimetry 

programs, it is unlikely that the information and internal exposure records provided in the worker 

files are complete. It is also likely that many worker internal exposures associated with highdose jobs were not monitored or documented. NIOSH’s missed dose approach and the default 

values should be complete and representative of all missed dose scenarios for personnel working 

at different INL facilities. At a minimum, NIOSH should evaluate the programmatic 

uncertainties associated with the missed dose values presented in Table 5.7-1, so that the 

recommended missed dose values would be truly claimant favorable. 

5.1.2.2 High-Risk (Dose) Jobs 

The DOE Tiger Team and DNFSB findings, made in the 1991 to 1994 time frame, suggest that 

there may be potentially significant missed doses due to deficient radiological protection work 

practices at different INL facilities. It is likely that the work practices at INL before this period 

were even worse, due to less sensitive equipment, less protective policy, programs, and 

procedures, and also less frequent and rigorous worker training. Instead of merely using 

inhalation dose defaults for worker missed dose from generic facility operational source terms, 

NIOSH should develop a list of high-risk jobs for different categories of workers (such as 

pipefitter, operator, RCT, machinist, mechanic, electrician, maintenance, yardman, etc.) at each 

INL facility-based on bioassay data, air sampling data, area survey data, and RWP data. High 

risk, in this context, means the potential for high dose exposure. Some of these high-risk (dose) 

jobs have been identified by past and present workers during the SC&A site interviews (see 

Attachment 3 of this report). 

It is generally agreed that the INTEC (previously ICPP) and its facilities are the most 

contaminated facilities at INL, with high potential for significant worker intake due to the fuel 
 

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processing and reprocessing operations and practices. NISOH should evaluate separately the 

missed dose for workers who had worked at ICPP facilities. 

5.1.2.3 Calibration of Internal Dosimetry Analytical and Monitoring Equipment 

The TBD does not provide any information on the calibration procedures, sensitivities, and 

standards of the internal dosimetry analytical equipment and monitoring instrumentation, 

including thyroid counters, whole-body counters, laboratory instruments, and continuous air 

monitoring systems (CAMs). The 1991 DOE Tiger Team findings show the deficiencies in these 

areas. NIOSH should evaluate the uncertainties and impacts on the internal dose assessment 

results associated with the deficient calibration programs at INL. 

5.1.2.4 Changes of Internal Dose Limits 

The TBD describes the historical changes of internal exposure control and monitoring programs 

at INL and touches on some of the internal dose limit changes. Currently, the INL policy is that 

all individuals who have the potential to receive a dose exceeding 100 mrem shall require 

monitoring. In the past, the INL policy was less restrictive and protective. For example, during 

the site expert interviews, a worker indicated that he often had positive nose swipes of 

100mrem/hr, but a bioassay or PEQ was not triggered. This type of inconsistent work practice 

was prevalent in the early years of the INL operation and may have led to significant missed 

dose to workers. 

NIOSH should evaluate the impacts of these dose limit changes over the operating history of 

INL to see whether there were missed doses in the early years when the radiation protection 

policy was less protective and inconsistently implemented. 

5.1.2.5 High-Fired Plutonium and Uranium Intakes 

Some INL facilities contained high-fired uranium or plutonium oxide produced by heating the 

material to approximately 1,000oC. High-fired oxide is more chemically stable than low-fired 

oxide, because the higher heat removes moisture and other impurities. The Rover facility at 

ICPP was used to reprocess graphite space reactor fuel, resulting in the formation of high-fired 

uranium oxide. It was identified as one of the 10 most significant highly enriched uranium 

safety concerns in the DOE complex. In addition, some high-fired plutonium oxides may have 

been shipped from Rocky Flats Plant and disposed of at INL. These oxides are commingled with 

radioactive wastes stored in drums and buried underground or above ground at the RWMC 

facility. 

Plutomium-238 is primarily an emitter of alpha particles, which are easily stopped by the 

materials encasing it, although it also emits some gamma rays and neutrons. The high-fired 

Pu-238 ceramic does not easily dissolve nor does it easily break into very small particles. In the 

case of plutonium, a urine sample would not show small intakes within two days, so INL staff 

developed a procedure for detecting fired plutonium oxide in fecal samples. Similar to recycled 

uranium, NIOSH should evaluate the lung dose for intake of high-fired uranium and plutonium 

oxide particulates (alveolar deposition). 
 

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5.1.2.6 Skin and Facial Contamination 

This TBD does not consider incidents with workers having skin contamination, facial 

contamination, and positive nasal swipes in the INL facilities. These kinds of problems would 

be compounded by the deficiencies in air sampling systems and ineffective respiratory protection 

programs, some of which have been discussed previously. Consequently, a bioassay may not be 

triggered for the workers. Guidance should be provided to a dose reconstructor to account for 

the missed dose due to the unaccounted uptake. 

5.1.2.7 Breathing Rate 

The NIOSH TBD (Section 4.2.1, pg. 10) uses an annual breathing rate of 2.4 x 103 m3/yr rather 

than the 2.88 x 103

 m3/yr (1.6 m3/hr for 1800 hr/yr) of ICRP 68 or the 8.0 x 103 m3/yr (RAC2002, 

pg. 38) of the NCRP. The TBD assumption appears less claimant favorable than the ICRP or 

NCRP assumptions. 

5.1.2.8 Non-Occupational Worker Elimination of DU Background at SMC 

The last paragraph of Section 5.6.1.1 of the internal dose TBD (pg. 29) on the SMC project to 

produce DU armor for tanks discusses natural background uranium baseline for radiation 

personnel working at the SMC facility. This section states, “Urine samples submitted by SMC 

nonradiation worker in 1987, 1994, and 1998 were assumed to represent nonoccupational 

elimination of the SMC worker population.” This study used urine sample results from 

nonradiation personnel working onsite at SMC in 1987, 1994, and 1998 to determine the 

nonoccupational component of uranium excretion for SMC radiation workers. The results 

ranged from 0.04 to 0.33 μg/L, with the average uranium concentration as 0.157 ± 0.109 μg/L at 

1 sigma uncertainty. The value of 0.16 μg/L was used and subtracted from each worker urine 

sample result prior to assessment of occupational internal dose. 

It is not sound to use a group of co-workers at the same facility as a reference, even though they 

are non-radiation workers. NIOSH should consider using non-INL individuals from the 

surrounding populations for the non-occupational elimination approach. The INL approach is 

not statistically conservative or technically valid due to the nature of the nonoccupational worker 

group used. The selected group of workers consisted of personnel collocated with the radiation 

workers. There is a high probability that these non-radiation workers had also been exposed to 

depleted uranium dusts or contaminated soils from the SMC grounds. NIOSH should, therefore, 

evaluate the validity of the value of 0.16 μg/L used as nonoccupational subtraction from urine 

results. 

In the teleconference, the NIOSH technical experts explained, “The SMC program was run 

independently from the other INL programs and considered only its own people for establishing 

background. NIOSH considered that, but using only the SMC population to determine 

background is consistent with the approach taken elsewhere on the site and at other sites (e.g., 

Fernald)” (see Attachment 2). The idea of consistency with the approach taken elsewhere on the 

site and at other sites (e.g., Fernald) is understandable and should be encouraged. However, in 

this case, data validity is more important than consistency. 
 

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In the site interview, the INL Dosimetry Department staff indicated that they had attempted to 

use non-INL workers from the surrounding populations, but the individuals they had contacted 

did not agree to sign a liability waiver form. That caused the change of plan to use onsite nonradiation workers. NIOSH should consider using uranium background data from other studies 

performed for the Idaho Falls area to compare with this value. 

A survey by the National Center for Health Statistics (NCHS) of CDC, National Health and 

Nutrition Examination Survey 1999–2000, shows the uranium concentration in urine for the U.S. 

population aged 20 and older ranging from 0.006 μg/L to 0.054 μg/L, with a geometric mean of 

0.007 μg/L. This is much lower than the INL background worker elimination value of 0.16 μg/L. 

In addition, NIOSH may consider not subtracting contributions from environmental background 

from the urine monitoring results. This approach is also used at Y-12 (Eckerman 1999). 

5.1.2.9 Unmonitored Workers 

As for the unmonitored (sometimes unbadged) workers, such as secretaries, warehouse workers, 

drivers, guards, and construction workers, the potential missed doses would be from inhaling 

resuspended contaminated soils and ingesting contaminated materials while eating in a 

contaminated, previously considered uncontaminated, area (such as office and cafeteria). 

NIOSH should evaluate these potential missed doses. 

5.1.2.10 Naval Reactor Facility Workers 

Release data from the Naval Reactor Facility was deemed classified until recently. As the 

internal dose TBD indicates, “some workers’ internal dose could have resulted from their support 

work at the NRF.” NIOSH should evaluate the potential missed dose at the NRF for these 

workers. 

5.1.2.11 Plutonium Monitoring 

The TBD does not provide any historical information on the plutonium analysis methods used at 

INL. Table 5.4-1 indicates a Pu-239/240 MDA starting in 1964. Table 5.3.2-1 indicates that 18 

Pu-239 urinalyses were performed in 1959. The site experts interviewed indicated that routine 

plutonium monitoring was not performed until the 1980s. There seems to be some confusion on 

this point. It is entirely possible that selective plutonium monitoring on workers was used at INL 

until 1980, but without this information, the dose reconstructors would not be able to assign 

missed internal dose due to plutonium intakes in the time period before 1980. NIOSH should 

provide information on plutonium monitoring in the TBD. 

5.1.3 SL-1 Accident Dose Reconstructions 

The TBD acknowledges that the highest sources of gamma and beta doses were the ICPP and 

SL-1. The criticality accident in the latter, which occurred at the end of 1960, caused many 

significant external exposures to workers involved in rescue and clean-up operations, but the INL 

TBDs treat this accident only sparsely. For instance, the Site Description TBD indicates the 

following: 
 

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The event released fission products (500,000 Ci in the building, and 1,100 Ci to 

the atmosphere) and created high-level radioactive contamination to 50 rad/hr 

around the ARA-II area. Initial recovery from the accident resulted in short term 

exposure exceeding 500 rad/hr to personnel in radiation fields. Extensive 

cleanup efforts followed, including complete dismantlement of the facility. 

(pp. 49–50) 

The Occupational Environmental Dose TBD states the following: 

The amount of the release and the path that the cloud traveled from the reactor 

building was carefully monitored and well documented. All radiological doses to 

personnel involved in the rescue and cleanup of the reactor building were 

carefully controlled and documented. The SL-1 accident did not affect any other 

INEEL facility with the effluent of radioactive material. (pg. 12) 

The Occupational External Dose TBD states the following: 

Experience following the SL-1 accident showed a wide variation of beta-togamma ratios and necessitated controlling both radiations rather than just the 

gamma. A set of as many as 18 badges could and in many cases was fastened on 

a belt around the worker to determine a beta:gamma ratio for each particular 

entry. (pg. 16) 

It is clear that the TBDs rely on the assumptions that (1) personnel dosimetry systems and dose 

assessment methodology following the accident were sufficient and accurate enough to calculate 

exposure doses to workers, and (2) the dosimetry record-keeping system was adequate to 

document dosimetry analysis and dose calculation results. However, these two assumptions are 

flawed, because dosimetry equipment of that era had limited capabilities to respond to an 

accident of such a large scale, and the dosimetry record-keeping system was still in its 

developmental phase. Additionally, there were close to 1,000 workers involved in that historical 

cleanup operation. It would be essential for NIOSH to identify high-risk (dose) jobs during the 

SL-1 rescue and cleanup process, and then provide specific claimant-favorable default dose 

values for each of these high-risk jobs. These high-risk jobs may include (1) fireman, (2) other 

first responder, (3) ambulance driver, (4) medical staff, (5) Hot Shop worker, (6) cleanup 

crewman, (7) machinist, (8) pipefitter, (9) crane operator, (10) RCT, and (11) guard. The missed 

dose could be composed of (1) missed beta dose, (2) missed gamma dose, and (3) missed internal 

dose. There may also be the potential of missed neutron dose for the rescue workers. 

Susan Stacy’s documentary book, Proving the Principle (Stacy 2001), devotes two chapters to 

describing the unfolding and the aftermath of the SL-1 incident. On page 147, it states the 

following: 

The SL-1 was a mess. It hadn’t been cleaned up at all. To clean it up, people had 

to make short trips inside and do limited tasks within a couple of minutes and then 

get out. Even though you suited up, those couple of minutes would expose you to 
 

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your quarterly dose of radiation, and you couldn’t go back in for three months or 

do any other work that could potentially expose you. Hundreds of people at GE, 

including those about to be transferred and many workers from other locations at 

the Site [and from Dugway], volunteered to take their quarterly radiation dose 

doing clean-up at the reactor. For many cleanup tasks, that was the only way of 

handling it.” I don’t remember anyone being particularly fearful of the risk. The 

time keepers were the HPs, who stood half-way down the stairway and banged on 

metal when someone’s vacuum-cleaning stint was over. By November, the 

passage of time and removal of debris had reduced radiation levels. 

It seems that there are many discrepancies in getting to the real facts about the SL-1 incident and 

also the rescue operations and the aftermath cleanup activities, since this occurred almost 45 

years ago. Over the years, there have been many complaints from INL workers and union 

workers about the lack of records and facts to assist in the dose reconstructions associated with 

this accident. The site expert, union worker, and retiree interviews at Idaho Falls raised similar 

complaints. The issue of “claimant favorable” for those claimants who participated in the rescue 

operation and/or in the aftermath cleanup activities must be addressed in a fair and reasonable 

way. 

Proving the Principle continues to state the following: 

Twenty-two of the people who had responded to the SL-1 alarm received 

radiation exposures in the range of 3 to 27 Roentgens total body exposure. Three 

of them received more than 25 R. The exposure guide that had been set up by 

IDO’s prior emergency plan allowed rescue personnel a 100 R dose to save a life 

and 25 R to save valuable property. 

Obviously, INL was not prepared and organized in 1961 to deal with such a unique and deadly 

event. The IDO Report on the Nuclear Incident at the SL-1 Reactor (IDO 1962) states the 

following: 

Following the incident, the routine service, of course, terminated; and it soon 

became apparent from the increased numbers of persons that were becoming 

involved with SL-1 rescue operations, etc., that a unique type of data processing 

system had to be instituted. The system would have to (1) maintain a radiation 

dosage record of those personnel brought in from other sites, (2) supply compiled 

information for appropriate officials and the person’s permanent work location 

indicating the radiation dose received while in the SL-1 Area, and (3) provide a 

total radiation dose tabulation for all officials concerned in order to prevent an 

exposure in excess of guide values... In order to prevent persons who were 

involved not only in daily duties at the SL-1 Area, but also in daily duties at other 

NRTS Site Facilities from receiving excessive exposure, each individual’s daily 

film badges, perhaps two or more, had to be processed and recorded. Next a 

report listing the individual’s name, date and location of exposure, and amounts 

of beta and gamma radiation received had to be tabulated. This created a record 

retrieval problem which would have proven extremely difficult, if not impossible, 
 

 

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without the central data processing system which maintains all radiation 

exposure records at the NRTS. (pg. 39) 

Following the incident, because personnel involved in the rescue operations were 

exposed to airborne radioactive materials, spot urine samples were collected 

from these people as soon as was possible. It was determined by gamma spectra 

analyses that the major portion of the radioactive material contained in the urine 

samples was iodine-131. Table 3.2 lists the sixteen persons who showed the 

highest urinary excretion of iodine-131 and the estimated thyroid dose that was 

calculated for each person.... In conjunction with the iodine-131 analysis, 

strontium-90 and cesium-137 analyses were performed. In all cases of the 

cesium-137 analyses, no activity greater than 15 milli-rem per year, infinity dose, 

was ascertained. Spot urine samples were collected from about 110 individuals 

at intervals during the first few days after the incident and were analyzed for 

strontium-90. Eight individuals whose urinary excretion of strontium-90 

continued beyond the 10th day then submitted 24-hour urine samples for strontium 

90 analyses. Table 3.3 lists the three persons who showed the highest strontium 

90 dose and the estimated strontium- 90 dose to the bone critical organ.... Six 

persons involved with the SL-1 Incident reported for whole body counting during 

the first week following the incident. Identification of iodine-131, bariumlanthanum-140, and cesium-137 was made by gamma spectra of all six o f these 

persons. However, the amount of iodine-131 made it impractical to do much 

further total body counting until most of the iodine-131 had left the body. 

(pp. 40–41) 

After the recovery of the first victim from the reactor operating room of the SL-1, 

a radiation survey was made of all personnel participating in the rescue 

operation. Eleven of the persons surveyed were found to be highly contaminated, 

many exceeding a radiation level of 10 R/hr on their extremities; and, as a result, 

they were directed by an IDO Health Physicist to be taken to the Gas Cooled 

Reactor Experiment (GCRE) Plant for decontamination. (pg. 42) 

According to the IDO report, during the rescue operation, the monitoring equipment used were 

not capable of reading the high dose rates in the SL-1 reactor building. 

At about the third or fourth step the Juno Radiation Detector which the MTR H.P. 

was carrying pegged at 25 R/hr. The Fire Captain and the MTR H.P., therefore, 

evacuated from the area stopping at the guard house. The total estimated time for 

this entry was one and one-half to two minutes. (pg. 23) 

At the top of the stairway, the door to the reactor operating room was found open. 

The Phillips H.P. held a high range Jordan Detector in the doorway and observed 

that the radiation detector pegged on the 500 R/hr scale. Next, with the Assistant 

Fire Chief holding a light in the doorway, in order to better illuminate the reactor 

room, the Phillips H.P. took a brief look into the room. Because either the light 

was dim or the mask’s eyepieces were fogged, the Phillips H.P. did not see very 
 

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much of the reactor room and none of the ceiling. He was able to observe some 

wreckage in the reactor room, but saw no personnel. They then hastened down 

the stairs and out the building via a door on the southeast side of the corridor 

leading from Building 602 to the reactor building. (pg. 25) 

At about 10:00 P.M. the MTR H.P. at the SL-1 was authorized to allow personnel 

to enter radiation fields greater than 200 R/hr by the enroute IDO Chief, Site 

Survey Branch; it was stipulated that Scott Air-Paks should be worn by all people 

involved in penetration operations. In view of this authorization, plans were 

made for another entry into the SL-1 Area. It was decided by the health physics 

staff at the scene that the planned entry should be halted when a radiation rate of 

500 R/hr was observed, because this was the maximum dose rate which could be 

read on the high range Jordan Radiation Detector..... At the top of the stairs a 

dose rate of 500 R/hr was observed, and it was estimated that a radiation field of 

500–700 R/hr existed about two feet inside the operating room. A quick 

observation made of the reactor room indicated one person positioned on the 

floor near the Motor Control Center (MCC) and moving, and another person 

positioned between a shield block and the reactor head. Radiation fields of 500 

to 600 R/hr were estimated to be near the MCC and greater than 1000 R/hr 

directly over the core. The two CEI personnel then departed from the reactor 

room to obtain assistance and equipment for a rescue. The entire entry and 

return operation took approximately three minutes. (pg. 26) 

Table 3.1 on page 44 of IDO 1962 summarizes the distribution of gamma radiation exposures to 

workers involved in the SL-1 rescue and reactor stabilization operations from January 5 to 

January 31, 1961. This table shows the following: 

(1) A total of 577 workers had participated in the operations 

(2) 413 workers received gamma doses ranging from 0 to 300 mR 

(3) 71 workers received gamma doses ranging from 300 mR to 900 mR 

(4) 66 workers received gamma doses ranging from 900 mR to 3 R 

(5) 18 workers received gamma doses ranging from 3 R to 12 R 

(6) 6 workers received gamma doses ranging from 12 R to 25 R 

(7) 3 workers received gamma doses over 25 R 

(8) The highest gamma dose received by any worker was 27 R 

There is no information given on the assumptions used in calculating these gamma doses. There 

are no data provided for internal doses, skin doses, and extremity doses. 

Tables 3.2 and 3.3 of the IDO report also summarize the thyroid doses and the bone critical 

organ doses estimated for some workers with airborne intakes. However, this report does not 

provide information for the gamma, thyroid, and critical doses to workers involved in the SL-1 

cleanup effort after January 31, 1961. 
 

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The 1963 technical paper (Horan 1963) entitled, The Health Physics Aspects of the SL-1 

Accident, provides a table summarizing the whole-body and thyroid doses received by 

emergency personnel, including the following categories: 

(1) AEC Health Physicist 

(2) Contractor Supervisor 

(3) Contractor Health Physicist 

(4) AEC Project Officer 

(5) Cadre Supervisor 

(6) AEC Physician 

(7) AEC Nurse 

(8) Support Patrolman 

(9) Support Health Physicist 

(10) Army Support 

As a “lesson learned,” the paper states the following: 

Disaster planning at the NRTS had been geared primarily to criticality type of 

maximum credible accidents involving the release of thousands of curies of fresh 

fission products or iodine-131 to the atmosphere. Many of the unique types of 

problems experienced or suggested by the SL-1 accident had not been considered: 

performing recovery operations in radiation fields of hundreds of R/hr, medical 

treatment and decontamination of highly contaminated survivors and casualties, 

performing field operations around the clock for an indefinite period of time. 

Supplies and equipment, with but a few exceptions, were adequate. One of the 

first lessons was that survey instruments with a maximum range of 500 R/hr are 

inadequate for emergency use. Instruments with a maximum range of 5000 R/hr 

should be available. Available health physics personnel were depleted due to 

overwork rather than overexposure before all equipment had been committed, 

and this despite the full support of 5 NRTS contractors and radiological 

assistance from 4 AEC or military organizations outside the State of Idaho. 

During the first 11 days, 81 health physicists were utilized in the field and over 

130 other personnel to provide them with field or laboratory support. Many 

individuals worked in excess of 120 hr per week, and yet sufficient personnel were 

not available to do the innumerable data collecting and research items which 

were desirable. The twenty-six man radiological detachment from the Army 

Chemical Warfare Center at Dugway, Utah, provided early and effective field 

support. (pp. 184–185) 

The INL Site Profile TBDs only provide minimal treatment of the SL-1 accident. But it is 

known that during the SL-1 excursion, 500,000 Ci of fission products were released in the 

reactor building and 1,100 Ci to the atmosphere. The accident also caused high-level 

contamination of up to 50 R/hr around the ARA-II area. There were more than 1,000 workers 

involved in different phases of the rescue and cleanup operations. It is imperative for NIOSH to 

provide specific guidance to dose reconstructors in calculating different missed doses for the 

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For the initial rescue operation (first phase), NIOSH should evaluate the missed occupational 

environmental doses to unmonitored and monitored workers (support workers), who arrived at 

the SL-1 facility but stayed outside the reactor building. These workers would be exposed to the 

fission products released in the surrounding atmosphere, and also the high-level radiation field of 

50 R/hr. It is plausible that these workers were exposed to short half-life fission products that 

were not considered in the TBDs. 

NIOSH should evaluate the missed occupational internal and external doses to rescue workers, 

such as firemen and health physicists, who either were involved in the discovery of the accident, 

made entries into the reactor building to retrieve the victims, or performed radiation surveys. 

There is no doubt that these workers would have received high level bodily contamination and 

inhalation intake of the fission products (500,000 Ci) released inside the building. The workers 

also had the potential of significant intake of uranium and its daughters and exposures to very 

high-level beta/gamma radiation fields that exceeded 1,000 R/hr. The other workers (such as the 

nurse or the doctors) who did not enter the reactor building, but made physical contact with the 

victims and equipment used to retrieve the victims, would have been contaminated as well. 

For the reactor dismantling operation (second phase), NISOH should evaluate the missed 

occupational environmental, internal, and external doses to workers, monitored or unmonitored, 

who were involved in retrieving the reactor core, fuel elements, and fuel fragments. For instance, 

the crane operators and maintenance workers would have been exposed to the high gamma 

radiation fields from the reactor core when the top cover of the reactor building was lifted, so 

that the reactor core could be retrieved and dropped into the shielded cask. NIOSH should also 

evaluate the missed occupational internal and external doses to workers who were involved in 

the examination and decontamination of the reactor core at the ICPP Hot Shop and Hot Cell. 

For the reactor building cleanup operation (third phase), NIOSH should evaluate the missed 

occupational environmental, internal, and external doses to workers, monitored or unmonitored, 

who were involved in entering or not entering the reactor building to perform cleanup activities. 

These workers would have been exposed to high-level beta/gamma radiation fields and subject to 

significant inhalation intakes of fission products and uranium. 

In view of all technical difficulties and equipment shortcomings in the SL-1 rescue and 

stabilization operations, NIOSH should evaluate the equipment used during the SL-1 incident 

and determine the uncertainties associated with the dose rate estimation, gamma dose assessment, 

airborne intake estimation (since there was no air sampling data), thyroid count, critical organ 

dose assessment, extremity dose assessment, and stay-time determination for various personnel 

categories. The results of the uncertainty evaluation would be helpful for the dose 

reconstructions. NIOSH should provide adequate information and explicit guidance in 

determining various types of missed dose for workers who participated in different phases of the 

SL-1 operations. 

In addition, NIOSH should also evaluate the possibility of missed neutron dose received by the 

workers during the rescue operations. Even though the Occupational External Dose TBD states 

that, “the high dose locations where most of the gamma and beta dose is received, such as the 

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the initial rescue operation of the SL-1 accident, there was no consideration given for neutron 

measurements of worker exposures when they were making entries into the reactor building 

where fuel rod fragments and debris were scattered all around the control room floors. 

5.1.4 Occupational External Dose Issues 

In the Occupational External Dose TBD, NIOSH presents thoroughly the history of the 

development of worker dose limits from the 1930s to the present. The TBD evaluates the 

external dosimetry programs, record-keeping practices, and different dosimetry systems (films, 

pocket ionization chambers, TLDs, neutron track emulsions, and albedo) used historically at INL. 

The TBD presents information (based on Reilly 1998) on the characteristics, calibration, and 

uncertainties associated with the dosimetry systems. The TBD also provides radiation fields 

measured at major INL facilities in 1998 with their relative biases. NIOSH summarizes all 

information associated with potential missed external doses in two separate categories (photon 

missed dose and neutron missed dose) for monitored workers in Tables 6B-1 and 6B-2. 

However, NIOSH did not evaluate facility field logs, RWPs, facility survey reports, incident 

reports, occurrence reports, contamination reports, PEQs, and ALARA records. The TBD does 

not provide evaluation of special exposure events, such as the SL-1 incident, to determine missed 

doses for monitored or unmonitored workers (e.g., nurses) during rescue, stabilization, recovery, 

and cleanup operations. NIOSH interviewed INL Dosimetry Department staff and requested 

individual dosimetry records and documentation. NIOSH also interviewed some workers and 

retirees as part of an outreach program after the TBD was done. 

As in other instances cited in this report, the authors of the Occupational External Dose TBD 

display confidence in the past and current INL radiological protection programs and 

implementation, the accuracy of the external dosimetry programs, including beta/gamma and 

neutron systems, and the dosimetry record-keeping systems. The TBD states the following: 

It was INEEL policy that personnel expected to receive any radiation dose or 

personnel whose work was centered at the site were assigned a radiation 

monitoring badge. (pg. 8) 

The INEEL dosimetry organization developed a set of basic administrative 

practices in 1951, which have changed somewhat as the technologies of ionizing 

radiation dosimetry and recordkeeping have changed. (pg. 9) 

When there has been a question about a dose value being assigned to an INEEL 

worker, a Personnel Exposure Questionnaire was normally initiated. (pg. 11) 

It is important to point out that many records and data pertinent to the worker claims are not 

included historically in the workers’ files, including personnel contamination reports, PEQs, 

RWPs, incident reports, and occurrence reports. 

NIOSH has included the following historical dosimetry standards and site administrative 

practices information in the TBD: 
 

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(1) The TBD identifies the way that INL reports external doses for workers: 

INEEL has reported doses as penetrating and nonpenetrating. The 

penetrating dose corresponds to the deep dose equivalent, and the 

nonpenetrating dose plus the penetrating dose corresponds to the shallow 

dose equivalent.” (TBD, pg. 6) 

(2) The TBD summaries the history of Federal dose limits for occupational workers: 

a. 1949 – 0.3 R per week or 15 R per year (recommended by NCRP-7) 

b. 1957 – (5N-18) rem as maximum allowable dose (NCRP) 

c. 1958 – 3 rem per quarter or 15 rem per year (AEC) 

d. 1960 – 3 rem per quarter or 12 rem per year (President Eisenhower) 

e. 1971 – 5 rem per year (NCRP 39) 

f. 1971 – defined the concept of deep and shallow dose equivalent indexes (ICRU) 

g. 1985 – introduced ambient dose equivalent, directional dose equivalent, 

individual dose equivalent penetrating, and individual dose equivalent superficial 

(ICRU) 

h. 1981 – required the monitoring threshold as 100 mrem effective dose equivalent 

(DOE 5480.1A) 

i. 1985 – specified the measurement of deep and shallow dose equivalents at depths 

of 10 mm and 0.07 mm, respectively (ICRU) 

j. 1990 – redefined the concept of dose equivalent to equivalent dose, quality factor 

to radiation weighting factor, and generated new factors (ICRP) 

The NIOSH approach in identifying the missed external doses at INL makes the following 

assumptions: 

(1) The INL site radiological protection programs, external dosimetry programs, recordkeeping systems, laboratory analytical programs, equipment calibration programs, quality 

control programs, and training programs were effective and dependable. This assumption 

that there was no administrative flaw in the implementation of these programs in 

protecting workers from external radiation was found to be unsubstantiated by the Tiger 

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(2) The information provided in the claimant files is complete for dose reconstruction. 

During site expert interviews, many past and current workers complained that there are 

dose records missing from their personal exposure file. 

(3) The dosimeters were accurate in measuring external worker dose within the projected 

uncertainties. This assumption is valid only if the dosimeters were calibrated correctly to 

measure the intended radiation fields, but there were documented deficiencies in INL’s 

calibration programs. 

(4) The workers had worn their dosimeters correctly in such a way that external dose was 

measured accurately. This assumption is not entirely valid, especially in varying high 

beta/gamma fields or in high-risk maintenance jobs, when personnel had to work in tight 

spaces very close to different sources. For example, sensitivity of dosimeters to low 

energy photons is angular-dependent. As a result, the dosimeters would under-measure 

the gamma doses. The TBD does not evaluate missed worker dose due to lack of multibadging or faulty badge placement in high, varying, or oblique beta or gamma fields. 

(5) There is no other scenario that would cause missed external dose to workers at INL. The 

approach to missed dose in this TBD precludes the potential for missed beta/gamma 

doses, due to inconsistent and less protective work practices in the early years. 

(6) NIOSH believes that the claimant-favorable approach to estimate missed external doses 

(both photon and neutron) is to use the following equation: 

(N x MRL/2), 

where N is the number of “zero” doses recorded. The MRL/2 approach is not claimant 

favorable. 

(7) There is no stand-alone missed beta dose for workers working at INL, such as low beta 

fields not measured by dosimeter (shallow dose with skin cancer concern) or very high 

beta fields measured as gamma (deep) dose. This is a questionable assumption (see 

Taulbee 2002). 

NIOSH believes its missed gamma and neutron dose values are claimant favorable, as the 

Occupational External Exposure TBD states the following: 

The missed dose for dosimeter results less than the MRL is particularly important 

for earlier years when MRLs were higher and dosimeter exchange was more 

frequent. One option to calculate the missed dose described in NIOSH (2002) is 

to estimate a claimant-favorable maximum potential missed dose where MRL/2 is 

multiplied by the number of zero dose results. (pg. 32) 

The use of MRL/2, or LOD/2 (Taulbee 2002), for calculating the missed dose values is 

questionable. This averaging approach may be adequate from a statistical standpoint for 

evaluating total missed dose for worker population, but at best, this approach is claimant-neutral 

for an individual. It is not claimant favorable for an individual worker, who may have received 

all his missed doses in the upper half above the average (MRL/2). It would be more appropriate 
 

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to use the value of MRL (instead of MRL/2), i.e., 95% (2 sigma). In addition, the use of such an 

equation for missed dose calculation requires the compilation of accurate MRL values and 

conservative uncertainties for all dosimeters and film badges used at INL. 

5.1.4.1 Beta/Gamma Dose Issues 

For photon missed dose, NIOSH presents, in Table 6B-1 of the External Dose TBD, maximum 

annual missed dose values for different films (e.g., 552 Dupont film, 558 Dupont film, 508 

Dupont film) and for different TLDs (LiF, LiF in Teflon, Harshaw Two-Chip, Panasonic FourChip), based on the MRLs for six different time periods; (1) 1951–1958, (2) 1958–1966, (3) 

1966–1974, (4) 1974–1975, (5) 1974–1985, and (6) 1986 to present. The missed dose values are 

provided for different dosimeter exchange frequencies (weekly, biweekly, monthly, quarterly, 

and semi-annually). 

For beta and gamma organ dose evaluations, the TBD provides the following instructions: 

For photons prior to 1981 the conversion factor from exposure to organ dose 

should be used. For 1981 and after, the conversion factor from deep dose 

equivalent to organ dose should be used. (TBD, pg. 34) 

However, the TBD provides no discussion, approach, information, method, or specific 

instruction for calculating missed beta dose for INL facilities. The TBD does provide some 

information and data on the beta radiation fields at INL in Section 6.3.4.2 (pp. 23–25). Table 6-5 

lists the beta dosimeter thicknesses and associated under-reporting with the following, somewhat 

cryptic, instruction. 

The fraction of beta dose measured shown in Table 6-5 is the average as 

described above. To determine the corrected beta dose, divide the nonpenetrating result from the dosimetry system by the values in the last column of 

Table 6-5. The reported dose will likely be somewhat higher than this because 

the calibration probably did not consider such a correction and reported the dose 

for the calibration exposure (TBD, pg. 25). 

This recommendation tries to capture beta radiation causation of skin cancer. The uncertainty 

for beta dose is estimated to be 50% at one sigma. This table is not included as part of the 

missed dose calculation or discussion. It is not clear whether NIOSH believes that there is no 

missed beta dose at all at INL facilities or the contribution from missed beta doses have already 

been included in the gamma missed dose calculation. In either case, NIOSH needs to clarify and 

justify its position in the TBD. 

As quoted earlier, INL reported doses as penetrating or non-penetrating. The penetrating dose 

corresponds to the deep dose equivalent, and the non-penetrating dose plus the penetrating dose 

corresponds to the shallow dose equivalent. It is not clear how a dose reconstructor should 

allocate the total missed photon dose calculated using the above equation; should this total 

missed dose be added to both the shallow and deep doses? 
 

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The photon missed dose defaults are to be applied to the dose reconstruction according to the 

following conditions: 

Missed photon dose for INEEL workers would occur where (1) there is no 

recorded dose because workers are not monitored or the dose is otherwise 

unavailable, or (2) a zero dose is recorded for the dosimeter systems for any 

response less than the site dose recording threshold (the MRL) (TBD, pg. 32). 

Some individuals who might occasionally visit site facilities but did little work 

with radiation, had badges at several different facilities. It is not appropriate to 

base missed doses on the multiple badges issued. Early on at INEEL, the badge 

change frequency was not the same for everyone. Workers with low probability of 

exposure were placed on a longer change cycle than those with more chance of 

exposure. Therefore, missed doses should be based on the actual change 

frequency for a person, and the frequency can be determined from the 

individual’s data package (TBD, pg. 9). 

NIOSH also gives instruction to DOE staff on what to provide a claimant for the dose 

reconstruction: 

DOE provided dosimetry information for a former INEEL worker, whose dose 

reconstruction is underway, should include a dose summary for the employment 

period and a copy of each weekly, monthly, quarterly, etc., form which will also 

show the work location, so the individual file could be several inches thick in hard 

copy. Each sheet is redacted so only the person of interest’s name and applicable 

information are visible. This file provides the recorded information as to the 

exchange period for the person of that time period. (pp. 9–10) 

NIOSH has included the following beta/gamma dosimetry information in the TBD to 

demonstrate the adequacy of its missed dose approach. However, this information shows 

significant weaknesses in several key areas. 

(1) The TBD indicates that the practice at INL is to subtract environmental radiation levels 

from the reported gamma doses. 

These badges were usually stored at the respective operational area 

entrance security gate for INEEL facilities. Control badges, which are 

used to subtract background radiation, have also been located there. This 

practice may lead to subtracting environmental radiation from site 

activities reducing the reported doses. Environmental radiation levels 

have been monitored for most of the life of the INEEL, originally with film 

badges and later with TLDs. Table 6-1 presents results of this monitoring 

at facility fence-line locations near the security gates which can be added 

to individual’s dose history or used for non-radiation workers working at 

the site. (TBD, pp. 8–9) 
 

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If the fence-line dosimeters were kept at the same place as the control dosimeters, then 

the following could happen: Say a worker at ARA had an annual dose of 500 mrem in 

1960, and that his dosimeters read a total of 726 mrem and 226 mrem was subtracted 

from the control dosimeters over the year. The TBD indicates that the workers dose will 

be calculated as 500 + 226 x (2000/8760) = 552 mrem. However, since the “real” 

background for the year is around 100 mrem, the workers dose should be 500 + 226 – 

100 mrem = 626 mrem. The NIOSH environmental dose subtraction method, therefore, 

is not claimant favorable. 

(2) The TBD provides some sample reporting forms used in early days in 1958 and 1959 

(Figures 6-1 and 6-2) and explains how the worker doses were recorded in these forms. 

The TBD, however, does not provide a history of the methods used at INL in recording 

different external doses in the worker files, such as penetrating, non-penetrating, shallow, 

deep, and neutron doses. NIOSH should provide instructions on how to reconcile the 

worker doses using these different methods, and also how to add the missed doses during 

the dose reconstruction. 

The TBD indicates that, “the personnel monitoring badges have always been considered 

more reliable than pencil dosimeters; so after the film badge results became available, the 

daily pencil readings were no longer considered doses of record.” In one case, NIOSH 

shows the large difference between the two readings: “In Figure 6-1, the pencil readings 

totaled 820 mR and the badges reported 0 mR for 18 badges” (TBD, pg. 11). NIOSH 

indicates these discarded pencil dosimeter reading values “can be recovered from the 

earliest forms for a worst-case estimate of dose” (TBD, pg. 11). Perhaps NIOSH should 

conduct a study on how extensive are these differences in pencil dosimeter and film 

badge doses. This is especially relevant based on the numerous reports from site experts 

that the film badge dosimeters they wore while working at the INL facilities were 

underestimating the actual external doses they received (see Attachment 3). 

(3) The TBD surveyed different beta/gamma personnel monitoring systems at the INL site. 

These systems include the following: 

a. 1951–1958: NRTS Self-Service System – The film badges used sensitive and 

insensitive DuPont type 552 films with 2 windows, one open and the other with a 

1 mm cadmium filter. The badges were processed weekly. The type 552 film has a 

threshold level of about 30 mR. DuPont type 558 films, with a threshold level of 

10 mR, were used at two reactor areas. 

b. 1958–1966: Multiple-Filter NRTS Film Badge – The film badges used DuPont 

type 508 films with three filters, including 1 mm cadmium, 0.013 mm silver, and 

0.5 mm aluminum. The badges were processed biweekly or monthly, with the 

exception of the high-dose areas where weekly processing continued. A minimum 

reporting level (MRL) of 10 mrem was used for both beta and gamma radiation. 

The TBD indicates that the densitometer was zeroed using the “blank film” – here 

taken to be a film that had not been used in the field. This would mean that this 
 

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“blank film” was used as the control film. This, in turn, would mean that the 

“fence-line dosimeter” or the “access-gate dosimeter” was not used to subtract the 

background from the occupational dose. This practice of using the blank film for 

zeroing the densitometer is claimant favorable, in that the film dosimetry service 

found a low background area for its film store. 

c. 1966–1974: Lithium Fluoride Teflon TLD System – This TLD system used a 

lithium fluoride disk with two 13 mm Teflon disks and a 1 mm cadmium filter. The 

badge could read 30 mR on a quarterly basis. “For normal monitoring, only the 

open window TLD was read and considered penetrating dose unless it read more 

than 125 mrem, in which case the shielded TLD was also read.” In 1968, the 

monitoring period was increased from 3 to 6 months. In 1972, some processing 

was increased to an annual basis. 

d. 1974–1986: Harshaw Two-Chip TLD System – This TLD system used two LiF 

TLD chips, with one covered by aluminum and the other by Mylar. The aluminumcovered chip provided penetrating doses at a tissue depth of nominally 1 cm. The 

beta dose was calculated from the difference between the two chips. The 

calculation was accurate only for the beta energy used in calibration. The practice 

was to read only the open window chip to determine if the non-penetrating dose 

was above 15 mrem and thus, reportable. If the threshold was exceeded, both chips 

would be read. 

e. 1986–present: Panasonic Four-Chip TLD System – This TLD system used three 

lithium borate and one calcium sulfate TLD elements with plastic and aluminum 

filtration. The MRL used was 15 mrem beta and gamma in 1986, 30 mrem beta and 

gamma between 1986 and 1989, and 15 mrem for gamma and 30 mrem for beta 

until 1993. Since 1993, the MRL is 10 mrem. 

(4)	 The TBD indicates the necessity of using multiple badges in some high beta radiation 

fields. 

Experience following the SL-1 accident showed a wide variation of betato-gamma ratios and necessitated controlling both radiations rather than 

just the gamma. A set of as many as 18 badges could and in many cases 

was fastened on a belt around the worker to determine a beta:gamma 

ratio for each particular entry. (TBD, pg. 16) 

However, the TBD does not provide adequate information on multiple badging or 

specific guidance on how to apply it to dose reconstructions. 

(5) The TBD describes the calibration programs at INL for beta and gamma radiation. In 

1959, INL used a radium source for gamma calibrations and uranium for beta calibrations. 

From 1975 on, a 2–5 Ci Sr-90/Y source was used to calibrate TLD badges. In 1981, a 

phantom was used in the calibration facility. In 1983, a natural uranium source was used 

for beta calibration. NIOSH lists uncertainties associated with beta and gamma 
 

 

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dosimeter calibrations historically at INL in Table 6-3. Several important associated 

observations include the following: 

Separation of penetrating dose from non-penetrating dose was an issue in 

1957 (Bennett 1957) and again in 1976 (Jenson 1976), particularly for 

ICPP where strong high energy beta fields were not unusual. 

(TBD, pg. 19) 

The 1991 Tiger Team Review of the INEEL site indicated that the INEEL 

contractor and the Idaho Operations Office using the same sources for 

calibration led to a conflict of interest or an advantage in DOELAP tests. 

(TBD, pg. 19) 

NIOSH did not evaluate the adequacy of these radium and uranium sources used for beta 

and gamma calibration. There is no information on whether the calibration sources 

address the full range of the beta and gamma fields at INL facilities. 

(6) The TBD discusses the radiation fields at the INL facilities based on an INL study (Reilly 

1998). It also shows that the percentage bias of the beta and gamma measurements lie 

within a +27% to -43% range. 

a. NIOSH identifies that these biases may have caused missed dose corresponding to 

low-energy photons at all INL facilities, especially at RWMC and at TRA 

radiography facility. 

b. NIOSH provides a list of selected beta and gamma energies for various INL 

facilities in Table 6-6 as IREP inputs for use in dose reconstructions. 

Given this information, NIOSH does not provide specific guidance in compensating for 

the biases determined. 

(7) The External Dose TBD discusses beta radiation fields and dosimetry: 

Beta radiation fields are usually associated with activation or fission 

product radioactivity that is outside of a container such as a spill or only 

lightly shielded or in hot cells. High beta fields were not unusual at the 

INTEC where large quantities of fission products exist. Pure high energy 

beta fields in some locations, particularly in the exhaust stream, have 

caused dosimetry problems because the badge shielding or instrument 

packages did not provide a full 10 mm tissue equivalent coverage and thus 

beta fields would be measured as gamma fields. (TBD, pg. 23) 

Beta field dosimetry became fairly accurate with the definition of 

DOELAP requirements in the early 1980s. Prior to that beta monitoring 

systems had various flaws, primarily in a detector too thick to give a good 

surface result or one that was covered with extra material. Calibration 
 

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was to high energy betas from either uranium or strontium. The dose from 

low energy betas will be missed altogether if the beta energy is not 

sufficient to penetrate the detector cover and will be underreported if the 

beta energy is not sufficient to penetrate the entire detector. (TBD, pg. 23) 

Table 6-5 provides the cover and detector thicknesses for the beta badges 

used at the INEEL... To determine the corrected beta dose, divide the nonpenetrating result from the dosimetry system by the values in the last 

column of Table 6-5. The reported dose will likely be somewhat higher 

than this because the calibration probably did not consider such a 

correction and reported the dose for the calibration exposure 

(TBD, pg. 25) 

Given these discussions, NIOSH does not provide any guidance in compensating for the 

potential missed beta doses. 

(8) The TBD discusses the uncertainties associated with gamma measurements, saying that, 

“A realistic estimate of total uncertainty for photon dosimetry is about 35% at one sigma” 

(pg. 34). The absence of information from inter-comparisons means that the evaluation 

of the photon dosimetry uncertainties is more subjective than objective. The fact that the 

largest dose contribution was from photons with E > 250 keV will reduce the uncertainty. 

It also should be mentioned that the uncertainty depends on the dose recorded. For lower 

doses, the uncertainty can be as high as 100%, while for higher doses, 35% should be a 

reasonable estimate. 

The TBD also states that 9 sets of 5 dosimeters were irradiated with combinations of 10, 

20, or 30 mrem of Ra gamma and U beta with measurability within ±12 mR with 95% 

confidence. The standard deviation should be about ±3 mR instead of ±12. This applies 

to the system introduced in 1958. 

(9) The TBD discusses the uncertainties associated with beta measurements: “The 

uncertainty for beta radiation is somewhat larger, an estimated 50% at one sigma” 

(pg. 34). A beta dosimetry system will only show reasonable uncertainties if the system 

is calibrated with the same beta energy spectrum as expected for the radioactive source 

that will be measured. This was not the case at INL, with the exception of the project for 

the fabrication of shielding using depleted uranium. This, alone, should make the 

uncertainty higher than 100%. More seriously, the dosimeter will only estimate dose 

rates for geometries similar to the calibration geometry. In reality, there was no 

comparison between the facility beta exposure and the calibration set-up. A thorax 

dosimeter can only give a response, such as “yes there was a beta dose” or “no there was 

no beta dose,” and sometimes not even this was possible. A wrist dosimeter might give a 

more consistent answer, if it can be shown that it was “looking” at the beta source. The 

statutory requirement of a claimant-favorable dose reconstruction process is achieved by 

(1) giving the benefit of doubt when there are unknowns, and (2) defining uncertainties 

for measured data and selecting the 99th percentile values of a Monte Carlo distribution. 
 

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5.1.4.1.1 Completeness and Quality of INL Gamma and Beta Dosimetry and Record Keeping 

Programs 

In addition to the 1991 Tiger Team and DNFSB findings listed in the above sections, there are 

several other findings specific to the INL dosimetry programs which call into question the 

completeness and quality of the dosimetry results and records. NIOSH should investigate these 

items and evaluate their impacts on the dose reconstructions. 

(1) The 1991 DOE-HQ Tiger Team found that the INL Nuclear Accident Dosimeters were 

deficient, because the overall gamma/neutron response was within ±40%, which did not 

satisfy the ±25% requirement specified in DOE 5480.11 (Vol. 2, pg. 4-175). 

(2) The 1991 DOE-HQ Tiger Team found that the INL personnel dosimetry program was in 

serious deficiency for several reasons (Vol. 2, pp. 4-176 and 4-177): 

a. Absence of a centralized, integrated dosimetry program 

b. Absence of field application and characterization studies to respond to the site 

dosimetry needs 

c. Absence of adequate technical staff for technical support and development 

functions to maintain a quality dosimetry program 

d. Absence of any calibration sources for daily functioning of the dosimetry 

processing and developing element correction factors and quality control 

e. Absence of reviews for personnel dosimetry and related procedures, 

characterization of site radiation conditions to assure the adequate coverage in the 

range of exposures and energies and type of radiation anticipated 

f. Absence of changes in algorithms necessary to correct dosimetry assessment for the 

intrusion of different radiation conditions 

(3) During the site expert interviews, past and current workers at INL facilities provided 

first-hand information about potential missed dose scenarios and deficiencies in 

personnel protection programs and dosimetry record keeping. Even though there is a 

sentiment that the INL radiological protection programs and the advancement of 

equipment and techniques have made dramatic improvements over the past two decades, 

the missed dose problems due to these deficiencies in the early years must be addressed 

in a fair and reasonable manner. 

5.1.4.1.2 Penetrating and Non-Penetrating Doses 

Historically, after the early 1950s, INL categorized gamma and beta doses as “penetrating” dose 

and “non-penetrating” dose, while in the early 1950s, beta-gamma (combined) dose was 

categorized as non-penetrating/penetrating. Neutron dose was maintained and reported 
 

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separately. Current dosimetry records report “shallow” and “deep” doses, with the notation that 

penetrating (deep) doses include neutron and gamma exposures. Records provided by INL in 

support of dose reconstruction include “shallow” doses, “deep” doses, and “extremities” doses. 

Neutron doses are provided separately, but are also included in the “deep” doses. 

The procedures and algorithms used in the film badge dosimetry service in the early days 

underestimated the Hp(10) dose, because the low-energy photons reaching the dosimeter were 

considered beta radiation. Surprisingly, the film service then added this beta dose (to the skin) to 

the "deep" dose, making that practice claimant favorable. However, the External Dose TBD also 

requires the dose reconstructor correctly to consider only the "deep dose" as Hp(10). However in 

doing so, the low-energy photon contribution to Hp(10) is lost. 

To be claimant favorable, the NRTS (an early name for INL) calculated the beta dose and the 

gamma dose, summed these two together, and recorded the result on the worker files as a wholebody dose. The current dose reconstruction process is applying correctly the gamma dose as the 

effective dose. The problem is that this gamma dose alone is not claimant favorable, as the 

information on dose due to low-energy gammas (E < 100 keV) has been lost. 

The TBD states that, “..it was agreed that all soft radiation would be considered as beta, 

and all the penetrating radiation as gamma for the purposes of reporting. This created a 

large safety factor in favor of the employee, as the soft gamma produced excessive 

darkening of the film emulsion...” 

This administrative decision made by the dosimetry system, which discounted the contribution of 

low-energy photons to Hp(10), was not in favor of the workers. On the contrary, low-energy 

gammas also cause absorbed doses in the ICRP 60 organs. Due to the high number of Compton 

scatters and the small mean free path for photons at low energies, the “effective dose per photon” 

at say 60 keV is greater than that for a 1.2 MeV photon. Low-energy beta radiation will not 

cause a dose to the ICRP 60 tissues organs, with the exception of the skin, but there may be a 

few cases when the dose due to very high-energy betas to the testes and breast are relevant. 

It is very difficult to estimate the reduction in the dose recorded for the workers resulting from 

the film service’s decision. For reasonably high-energy photon fields, such as for reactor work 

or reprocessing plants, a 15%–30% underestimation of the effective dose may be possible. 

5.1.4.1.3 Correction for Beta Dose 

For the film dosimetry system and the Harshaw two-chip TLD system after it, it seems that INL 

considered only two irradiation possibilities; high-energy photons and beta radiation. However, 

OCAS IG-001, the external dose reconstruction guideline (Item 2.1.1.2), states the following: 

Within the NIOSH-IREP probability of causation program, there are three photon 

energy bands; 1) below 30 keV, 2) 30 to 250 keV, and 3) above 250 keV. 

Therefore, some separation of the dose from each energy band is required. 
 

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At most of the larger facilities, multi-shielded film badges or multi-element TLDs 

have been used since the mid 1960s. Since only three energy bands are used in 

the probability of causation calculations, the differences between various filter 

doses can provide insight into the gross energy distribution at the facility. To the 

extent possible, these differences should be used to estimate the relative energy 

distributions in earlier years when only two element film badges were used. If 

individual energy distribution information is not available for two element film 

badges, the open window dose should be used as a claimant friendly estimate of 

the 30 to 250 keV dose. 

When an estimate (of the energy spectrum) based on radionuclide inventory (the 

type of radionuclide) is conducted, some consideration should be given to the 

degree of Compton scattering that would contribute to the 30 to 250 MeV energy 

range. 

Considering the entire open window dose to be due to beta radiation is not claimant favorable. 

The main problem is that the filtered part of the dosimeter will not see the low-energy photon 

radiation (below 100 keV), and will under-report the dose for photons between 100 keV and 

approximately 200 keV. When it is assumed that a two-element film dosimeter is looking at 

photon radiation only, the following steps are made to calculate the dose: 

(1) If the net optical density (NOD) under the two filters is about the same, then the dose is 

due to high-energy photons (E> 250 keV) 

(2) If the NOD under the open window is higher than that under the silver filter, then an 

estimate of the energy of the photon radiation is made and the dose is estimated using the 

response curve of the film; a curve similar to Figure 6-7 (pg. 15) 

In the case of INL, this procedure (or a similar procedure) was not followed. The INL procedure 

considers that the open window measures the beta dose only, and the filtered part (also called the 

shielded part in the INL documents) of the dosimeter measures only the high-energy photons. 

The question now is how to estimate the low-energy photon dose from the available information. 

As the “beta doses” were estimated using the uranium daughter beta source for calibration, it is 

not possible to estimate the low-energy photon dose directly from the estimated beta dose. In 

fact, most of these “beta doses” are really low-energy photon doses. 

As it seems that the original optical density information was recorded, it would be possible to 

calculate the low-dose information for a few cases. In addition, since gamma spectroscopy 

information for some facilities is available, it might also be possible to check whether there was 

a significant presence of photons in the 30 < E < 100 keV range. If it can be shown from a few 

cases that a significant part of the photon dose was “missed,” then a global “correction factor” 

should be multiplied to the individual deep dose estimate, much in the same way the factor of 

1.119 was applied in the SRS TBD site profile. 
 

 

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5.1.4.1.4 Angular Dependence Correction Factor for Gamma Dose 

NIOSH should provide an angular dependence (anatomic geometry) correction factor for 

external photon doses. This correction factor would be used to account for the bias of a 

dosimeter worn at the neck level and the higher doses received by tissues/organs below the waist. 

The angular dependence of the sensitivity of the dosimeter is most pronounced at low-photon 

energies (SC&A 2005, pp. 143–145). 

5.1.4.1.5 Restating Beta Dose as Gamma Dose 

The TBD indicates that if individual energy distribution information is not available for twoelement film badges, the open window dose should be used as a claimant-favorable estimate of 

the 30 to 250 keV dose. Most of the time, while performing normal work as a plant operator or 

supervisor, the workers film badge or TLD will not get close enough (less than 1 meter) to 

surfaces emitting beta radiation to detect a beta dose. A worker (such as a pipefitter or welder), 

however, may for short intervals bring his dosimeter closer to a surface contaminated with beta 

emitters. Therefore, it is not claimant favorable to state that the entire dose measured in the open 

window is due to the beta dose. It is more correct to say that it is from photons in the range of 30 

keV < E < 250 keV. 

5.1.4.1.6 Photon Spectrum Split 

As the External Dose TBD indicates, NIOSH is assuming a 25%/75% split between low- and 

high-energy bands for the photon spectrum; SC&A recommends a more claimant-favorable split 

of 50%/50%, similar to that used in the SRS TBD. Whatever the split of the photon spectrum, 

the question is; How will the dose for the energy region 30 keV < E < 250 keV be calculated? 

As no photon open window dose was recorded, the dose in this energy range cannot be directly 

estimated. The difficulty of assigning correct doses can be illustrated by considering two 

extreme examples: 

(1) A worker receives a dose of 200 mR from a 60 keV source. The open window 

information would be turned into a beta dose to the skin. The filtered window would 

record little or no dose, say 30 mR. 

(2) A worker receives a dose of 200 mR from a 662 keV source. The photon energy is 

degraded by Compton scattering, so that in reality he receives 100 mR from photons in 

the energy range 30 keV < E < 250 keV and 100 mR from photons in the energy range 

E > 250 keV. The open window dose is turned into a beta dose to the skin. The filtered 

window dose registers around 150 mR, as some of the lower energy photons are filtered 

out. 

Considering that the dose reconstructor does not know the photon energy spectrum the worker 

was exposed to, what dose value would be assigned for the 30 keV < E < 250 keV energy region, 

and what dose would be assigned for the E > 250 keV region? 
 

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5.1.4.1.7 Immersion Doses 

Immersion in a cloud of beta emitters may happen periodically at INL facilities and lead to 

important consequences for internal dosimetry. The internal dose to the individual may be 

higher than that measured by a personnel dosimeter by one or more orders of magnitude. A 

finger dosimeter would not provide the required exposure information, as it would be filtered and 

therefore could not be used to estimate beta dose. A wrist dosimeter would give a better estimate, 

however, if it were located on the side of the wrist facing the beta source. 

It was said that during the teleconference with NIOSH experts that the film or TLD monitors 

would record the dose received due to immersion in a cloud of gamma emitters. The dosimetry 

system is calibrated for AP irradiations. The dose recorded on a dosimeter due to a semi-infinite 

cloud irradiation would be approximately half of the actual dose received. NIOSH should, 

therefore, consider a weighting factor of 2 for immersion dose. 

5.1.4.1.8 High-Risk (Dose) Jobs 

The TBD indicates that there are facilities at INL (e.g., INTEC) where high beta fields exist. 

NIOSH should develop a list of high-risk (dose) jobs and provide corresponding beta/gamma 

dose rates and worker job doses. This information will be helpful for dose reconstructions for 

personnel who had worked in such jobs and areas. Working in areas where there are fragments 

or “hot particles” of fission products, for example, during the cleanup of a reactor destruction 

experiment or SL-1 accident, may lead to the deposition of hot particles with high beta dose rates 

(above 50 rad hr-1) on the clothing and, possibly, directly on the skin of the face or hands. The 

beta radiation emitted from these hot particles will not be detected by the film or TLD dosimeters. 

For workers at fuel element or reactor cleanup operations, for small localized areas, the beta dose 

could be as high as 100–1,000 rads when calculated over a working day. For claimants with skin 

cancer, location and job-specific information should be taken into account. 

5.1.4.1.9 Extremity Dose 

NISOH should evaluate the potential for missed extremity dose for workers (such as operators, 

pipefitters, machinists, mechanics, or instrument technicians) working in facilities, where highly 

contaminated equipment, piping, instruments, valves, and systems resulted in exposures in 

confined spaces to hands. This is especially true for facilities, such as reactors, fuel processing 

cells, and waste storage systems. 

5.1.4.1.10 Discrepancies Between PIC and Film Reading 

Many facilities at INL contained areas with high beta radiation rates. However, as the distance 

between a worker and a planar beta-gamma source increases, the beta dose rate will fall off more 

quickly than the gamma dose rate. Therefore, at a typical working distance from contaminated 

surfaces, the beta dose rate will be lower than the gamma dose rate. 
 

 

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PIC dosimeters in the early years were fairly rugged, heavy devices, with steel cylindrical walls, 

and, therefore would record little contribution from beta radiation (perhaps a factor of 100–1,000 

times lower than an open window film dosimeter would record). Additionally, PICs are known 

to have a larger uncertainty than films or TLDs, and are relatively fragile and easily disturbed. 

For example, when a PIC is dropped on the ground or otherwise shocked, the reading can shift 

anywhere, even off the scale. In addition, some PICs are "leaky" and will discharge quicker 

(show higher doses) than they should. 

The INL workers interviewed have a confidence problem with the film dosimeter readings (see 

site expert interview notes in Attachment 3). However, if the PIC readings and the film readings 

are compared over a period of 1 year, and it is seen that the average comparison of PIC and film 

doses differs by more than say 30%–50%, this is evidence that the "deep dose" as reported by the 

film dosimetry service may have underestimated the Hp(10) dose. 

Workers consistently had access to their PIC results, and management periodically reported their 

film badge results to them. In later years, personnel were issued an annual radiation exposure 

report. Hence, personnel were able to make their own comparisons between PIC and film badge 

readings. Interviewed site experts expressed concern regarding the underestimation of actual 

exposure by film badges. For example, a site expert recalled that he received 500 mR on his PIC, 

yet his film badge result was 100 mrem for the same period of time. When radiological control 

personnel were asked to explain the difference between the two readings, they indicated that the 

PICs were likely responding to the beta exposure. The β to γ ratios can vary from 1 to 1 to as 

high as 25 to 1 depending on the facility. The TBD should compare PIC verses film badge data 

(i.e., shallow and deep) and ensure that all the dose has been captured by the film badge. It is 

important to note that some PICs were worn for only the length of the job, so the discrepancy 

between readings of the two dosimeter systems cannot be explained by drifting. 

An accurate estimate of the total "missed dose” may not be possible, but a rough idea of the 

"photon missed dose" could be obtained for each facility by the following: 

(1) Calculating the average dose and collective dose measured by film dosimetry shortly 

before the change to TLD dosimetry 

(2) Calculating the average dose and collective dose measured by TLD dosimetry shortly 

after the change from film dosimetry 

(3) Comparing the two and possibly use the difference as a direct estimate of the “film lost 

photon dose” 

5.1.4.1.11 Minimum Detection Limit 

Determination of the minimum detection limit (MDL) takes the dosimetry system uncertainty 

into consideration. The mathematics of this process is discussed in the External Dose TBD. 

When using state-of-the-art dosimetry systems today, with a 20 mrem (0.02 mSv) MDL, there 

will be a number of "false positive" results. As stated in the TBD (pp. 13–15), the necessary 

information to determine the MDL is not given in the NRTS file, but a value of 30 mrem chosen 
 

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by INL for Type 552 film would have been an optimistic MDL. A value of 40 mrem (as adopted 

by SRS) would be more realistic for high-energy photons recorded many years ago. Unless 

actual reports on tests of the MDLs of the INL dosimetry systems are available and are favorably 

evaluated, the more conservative SRS values could be used as surrogate MDLs for INL. 

The following is based on Determining the Lower limit of Detection for Personnel Dosimetry 

Systems. (Health Phys., 62(1): 2-9;1992.) 

Recording a result of a dosimeter as a dose when the dosimeter was not irradiated is known as a 

Type I error. Most laboratories set their acceptance of a Type I error at 5% when calculating the 

MDL for a given assay. That is, for any value that is greater than or equal to the MDL, there is a 

95% confidence level that it represents the detection of a true dose. The MDL (LD) for the film 

dosimetry system may be derived from the following equations (see figure below). 

LC = kα σo + Bo 

LD = LC + kβ σD 

If kα = kβ = k, and for low values of σμ (note: see equation 14 of the above paper), then LD may 

be considered to be approximately (for a 95% confidence level) 

LD = 3.4 σo 

Unfortunately, the values cited for the 10, 20, and 30 mrem irradiations may not be used to 

estimate σo. For this calculation, approximately 20 or more non-irradiated monitors should be 

evaluated. The standard deviation of the optical densities may be used and then converted to a 

standard deviation in dose by multiplying by the optical density-to-dose coefficient for highenergy photons. This coefficient, from figure 1 of the report, is about 10/0.02 = 500 mrem per 

optical density unit. The LD may then be calculated by multiplying the standard deviation in 

dose by 3.4. 
 

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The TBD mentions only the uncertainties due to the densitometer: ± 0.02 optical density units. 

The uncertainties due to the densitometer reading are the smallest and best known of all the 

uncertainties involved in the process. The largest uncertainty is the variation of the optical 

density due to the film development process. Very strict quality control measures help to reduce 

these uncertainties, especially variations between film batches. For example, the temperature of 

the film baths in one laboratory is controlled to 20°C ± 0.3°C, and even so, due to the continued 

use of the developing chemicals, the uncertainty in the optical density at 0.06 is ± 0.02. The 

unevenness of the film emulsion adds a further 0.01 optical density units to the uncertainty at this 

level. 

These uncertainties do not take into consideration the use of the dosimeter in the field, with the 

additional associated uncertainties of energy and angular dependence, temperature and humidity 

considerations, etc. Looking at contemporary and modern film dosimetry systems, for highenergy photons, the MDL for the INL dosimetry system could not have been less than 30 mrem, 

and 40 mrem (as used by the SRS profile) would be considered more likely and claimant 

favorable. 

5.1.4.1.12 Minimum Reporting Level 

NIOSH uses the detection threshold levels of the film badges and the TLDs as the MRL values 

for the missed dose calculation. However, the information NIOSH used to determine these 

MRLs is not complete and sometimes not supported. For instance, there is no information 

provided for the threshold levels for the original LiF Teflon TLDs and the Atlas TLDs. For the 

Panasonic TLDs, the information is quite confusing for dose reconstruction. NIOSH should reexamine the issue and provide more supporting information and a clearer explanation. In 

addition, the TBD does not provide any information concerning the uncertainties associated with 

these threshold levels. NIOSH should provide guidance to the dose reconstructors in assessing 

MRL uncertainties to ensure that the missed dose calculations are claimant favorable. 

5.1.4.2 Neutron Dose Issues 

For assessing missed neutron dose, the TBD adopts the sentiment that the INL neutron protection 

programs, dosimetry program, and record-keeping systems were reasonably complete and 

effective. This is shown in the following selection of quotations from the External Dose TBD: 

Most INEEL workers are not exposed to neutrons and so are not badged to 

measure neutrons. (TBD, pg. 25) 

Sources of neutron exposure include neutron sources at the instrument calibration 

laboratories and 14-MeV neutron generators used to characterize waste. For 

these spectra, the NTA works reasonably well (TBD, pg. 25) 

Individuals who have the potential to receive neutron dose currently wear albedo 

badges, and experience has shown that most do not receive significant doses.” 

(TBD, pg. 25) 
 

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Most of the reactors built at the INEEL had no beam ports. Thus the neutrons 

were generally well contained away from the workplace. The reactor core 

environment is characterized not only by high neutron levels, but also by very 

high gamma levels. The gamma shielding is often water and concrete which are 

also very good neutron shields. The neutron fields in the energy spectrum for 

reactors and lower will be attenuated much more quickly in concrete or water 

than will the gamma fields. This is not true for lead or iron, but these are usually 

not used as gamma shields where neutrons also exist. Thus neutron fields are 

generally not a problem at an enclosed reactor. [Emphasis added] (TBD, pg. 25) 

In Table 6B-2 if the TBD, NIOSH provides maximum annual missed dose values for different 

neutron dosimeter types (NTA and TLD) based on laboratory MRLs for three different time 

periods; (1) 1951–1958, (2) 1958–1976, and (3) 1976 to present. NIOSH believes these missed 

dose values are claimant favorable. The TBD provides these instructions for using the missed 

dose table: 

To calculate the missed dose, the reconstructor must first determine if the person 

worked near neutrons and which category of neutrons. This can best be done by 

looking for the work location and whether a worker or others in the badge 

reporting group were assigned any neutron dose equivalent. The work location 

code for TRA where the MTR operated is 4 (also 40 to 45). If no neutron dose 

was assigned to him/her or to co-workers for several months, the dose 

reconstructor should assume that the person was not exposed to neutrons so no 

neutron dose would be missed. 

If a worker was likely exposed to neutrons, the reconstructor should assign 

missed neutron dose equivalent using Table 6B-3 for the times when workers did 

not have reported neutron dose. For the period when NTA film was used, the 

dose should be multiplied by 1.25 for all facilities except the MTR experimental 

floor and by 2 for the MTR experimental floor when the MTR was operating 

between 1953 and 1970. Then the dose equivalent is apportioned into the IREP 

groups using Table 6B-3. (pg. 33) 

The TBD gives an example (the only example in the 6 TBDs) to demonstrate how the missed 

neutron dose for a worker at MTR can be calculated. This represents a commendable strength in 

the TBD that would have been welcome elsewhere as well. 

For example, if in 1955 a person was an experimenter at the MTR, and 7 of the 

weekly badges recorded a total of 185 mrem neutron dose equivalent, the missed 

dose would be 315 mrem [(52-7)×14÷2] so the total dose by the badges would be 

500 mrem. Because the badge only sees about one-half the MTR neutron dose 

equivalent (from Section 6.3.4.3.1), the total dose equivalent is 1 rem. To convert 

the 1 rem received from neutrons on the MTR experimental floor to equivalent 

dose, multiply the total dose equivalent by the last column of Table 6B-3 to get 

200 mrem to the <10-keV group, 60 mrem to the 10- to 100-keV group, 700 mrem 
 

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to the 0.1- to 2-MeV group, and 280 mrem to the above 2 MeV group for a total 

equivalent dose of 1.24 rem (pg. 33). 

For neutron organ dose evaluation, the TBD provides the following instruction: 

The calculated neutron dose equivalent to organ dose in each energy group 

should be multiplied by the conversion factors from ambient dose equivalent to 

organ dose for AP irradiation from Appendix B of NIOSH 2002 (pg. 34). 

NIOSH has included the following neutron dosimetry information in the TBD: 

(1) The TBD indicates that neutron monitoring was not automatic for workers at INL 

facilities. As stated on page 17, “Kodak nuclear track emulsion-Type A (NTA) was used 

for neutron monitoring when the field radiation protection staff requested it. NTA 

responds to neutrons with energies above 500 to 800 keV, for which the proton recoil 

tracks leave enough energy to expose at least three (four in some references) grains of 

emulsion.” [Emphasis added] 

(2) This type of neutron detector responds to neutrons with energies above 500 to 800 keV. 

The TBD assigns MRL values used for neutron track emulsions Type-A (NTA) at INL 

for two different time periods: 

a. Before 1958, MRL = 14 mrem 

b. After 1959, MRL = 20 mrem 

NIOSH’s approach in the determination of these two MRLs is not sound, as they are based 

on the lowest recorded neutron doses in several dose cards. 

(3)	 The TBD identifies that there were missed neutron doses below the NTA energy 

threshold of 0.5 to 0.8 MeV, particularly at plutonium facilities. Hankins (TLD) 

dosimeters were used at INL to capture thermal neutrons. The TBD provides INL facility 

neutron correction factors (FNCF) for 1981. 

(4) The TBD states that assuming INL changed from NTA to Hankins albedo dosimeters in 

1976 is claimant favorable. 

(5) The TBD describes that the calibration of the NTA was done with a 30-Ci PoBe source in 

1958. The uncertainties assigned were at the 90% confidence level. In order to 

compensate for the energy limitations of these emulsions, neutron pencil dosimeters were 

also used. In 1982, an AmBe source was used. In 1993, INL changed to an unmoderated 

Cf-252 source for calibration. The TBD lists in Table 6-4 the sources of laboratory bias 

in neutron calibration historically at INL. 

Alpha particles from the americium or polonium interact in the reaction 

9Be(α,n)12C, and generate a broad spectrum of neutrons up to about 11 

MeV (mean energy about 5 MeV). [Emphasis added.] (TBD, pp. 19–20) 
 

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(6)	 The TBD evaluates neutron radiation fields and measured neutron doses at INL facilities. 

1n 1979, the TBD indicates that five people received neutron doses between 0.5 and 

1 rem, and 79 received measurable doses below 0.5 rem. The TBD also represents 

(Figure 6-11) that only 54 out of 1,461 neutron dosimeters measured reportable doses in 

1995, and that only 6 of the 54 reportable doses were above 35 mrem. The MRL for 

neutron TLDs was taken as 15 mrem. 

(7) The TBD presents in Figure 6-12 the results of a relative bias study on neutron field 

measurements. 

(8) The TBD identifies the sources of neutron exposure at INL, including the instrument 

calibration laboratories, neutron generators, and reactors. However, the major 

contributors are the Material Test Reactor (MTR) and the Test Area North (TAN) Fuel 

Storage Casks: 

a. MTR – The MTR, which operated from 1952 to 1970, had beam ports and neutron 

beams extending onto a research floor. 

The exception to the above discussion is the MTR [Material Test 

Reactor], which operated from 1952 to 1970 and had beam ports 

and neutron beams extending onto a research floor. Also in this 

category are the Zero Power Physics Reactor (ZPPR) and 

Transient Reactor Test (TREAT), both at Argonne West. Some 

neutron surveys of the MTR experimental floor have been 

recovered (Sommers 1959, 1962; Hankins 1961), but these 

individually do not provide all components of the radiation field. 

(TBD, pg. 27) 

The supplement to this TBD describes in detail a 1961 neutron field study at MTR 

by Hankins. The TBD acknowledges that MTR personnel would have missed dose 

for neutrons with energies below 0.5 to 0.8 MeV. To correct for this missed dose 

for exposures on the MTR experiment floor, the TBD recommends multiplying the 

NTA results by 2 ±0.2 for a Monte Carlo dose reconstruction or by 3 for the less 

accurate worst-case dose reconstruction. The TBD also concludes that the total 

neutron dose equivalent is 0.58 ± 0.48 of the gamma dose equivalent on the MTR 

experiment floor. NIOSH believes many workers wearing NTA film badges would 

receive gamma dose at locations other than on the MTR experiment floor while the 

reactor was operating. For example, an RCT may be covering jobs with only beta 

gamma fields, and a craftsperson may be servicing pumps carrying radioactive 

water and not receive any neutron dose. The TBD recommends using a lower 

multiplying factor of gamma dose equivalent; 1.6 or 2.1. This approach is not 

claimant favorable. 

b. TAN Fuel Storage Casks – Some fuel storage casks were located on a storage pad at 

TAN. The dose rates were 25–30-mrem/hr gamma and 40-mrem/hr neutron. The 
 

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metal casks attenuate the gamma radiation, but not the neutron field, by a 

significant amount. The casks were moved to the Warm Shop for tests, where they 

remained for about 2 weeks before being moved back out to the pad. Neutron 

levels were discovered in the offices by an area TLD albedo system. Six people 

were not wearing albedo neutron badges. The estimated dose equivalent for full 

time occupancy is less than 50 mrem for each cask evolution taking into account 

attenuation by distance between the casks and office and by shielding from building 

concrete. 

(9) The TBD summaries the typical neutron personnel dosimeter parameters important to 

Hp(10) performance for the dose reconstruction in Table 6-7. NIOSH identifies the most 

important parameter as the difference between calibration and workplace neutron energy 

spectra. 

(10) Table 6B-3 of the TBD lists the locations at the INL facilities where it believes exposure 

to neutron radiation is credible. These locations include instrument calibration 

laboratories, waste characterization facilities (RWMC, SWEPP, 14-MeV neutron 

generator), neutron source research laboratories, and reactors (MTR, ZPPR, and TREAT). 

(11) The TBD presents the neutron weighting factors (function of energy) used to replace the 

neutron quality factors. With these weighting factors, INL corrected the dose equivalent 

to a newer dose quantity: 

Table 6-8 lists the calculated fractions of dose equivalent in the IREP 

energy groups and the conversion factors from dose equivalent to 

equivalent dose for INEEL spectra. The ratios of average radiation 

weighting factor to average quality factor for the IREP energy groups 

have some variation, particularly for the 10-100 keV group where the 

energy dependence of the fluence is radically different for the fission and 

14 MeV source than for the reactor spectrum. The lower part of the table 

lists the recommended default values for the dose equivalent fractions and 

quality factor corrections. (TBD, pg. 31) 

(12) The TBD estimates the uncertainties associated with neutron measurements as follows: 

The total uncertainty for neutrons is probably larger at about 60% at one 

sigma. The cause of the greatest uncertainty for neutrons is the variation 

of dose caused by an organ’s position in the body. For 1 MeV neutrons, 

the dose facing the source is about a factor of 1000 higher than the dose 

on the back side of a 30 cm diameter sphere of tissue- like material. In a 

work environment, the direction of the neutrons may be unknown, but it is 

often from many directions which reduces the impact of this uncertainty 

driver. (TBD, pg. 34) 

(13) NIOSH provides a TIB supplement to this TBD. This TIB reanalyzes the neutron field 

measurements performed by Dale Hankins using Bonner Balls in 32 locations around 
 

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MTR at INL in 1961. The Hankins data were reanalyzed using detector responses 

calculated by Hertel and Davidson for 171 energy groups from thermal to 17.3 MeV. 

Neutron dose equivalent (mrem/hr) values were determined for these field measurements 

and are listed in Table 1 and Table 2. The comparisons (ratios) of the different 

components of the neutron dose equivalent rate are provided in Table 3. The NTA 

neutron dosimeters used at MTR respond only to neutrons above 0.5 to 0.8 MeV. The 

TIB also presents the dose equivalent rate, the fraction of the dose equivalent, the ratios 

of the neutron ambient dose equivalent and the neutron dose equivalent, and the 

corresponding NTA response in each of the IREP energy regions in Table 4. 

In summary, the TBD draws the following conclusions: 

(1) NTA dosimeters were used to monitor workers for neutron exposure between 1951–1976 

(2) Monitoring of workers was performed only at the request of radiation protection staff 

(3) Minimum reportable levels (MRLs) for neutron exposures were defined at 14 mrem
 

(1951–1958) and 20 mrem (1959–1976) 

(4) Before 1982, calibration of NTA dosimeters was performed with PoBe source with a 

broad neutron energy range of up to 11 MeV with a mean energy of 5 MeV 

(5) Principal sources of neutron exposure to INL workers were the instrument calibration 

laboratories and 14 MeV neutron generators. Because most of the INL reactor fuel cores 

were “shielded” by water and concrete and had no beam ports, neutron fields created by 

reactors were “. . . generally not a problem.” 

5.1.4.2.1 Minimum Reporting Level 

NIOSH did not use the detection threshold levels of the NTA emulsion and the albedo TLD 

dosimeters as the MRL values for the missed neutron dose calculation. Instead, NIOSH 

reviewed a limited number of neutron dose data sheets to identify the lowest dose assigned to the 

worker. This lowest dose was then used as the MRL value. For instance, NIOSH reviewed one 

data sheet from March 1958 and determined that the MRL, for the period between 1951 and 

1958, was 14 mrem, because that was the lowest reading of the 10 neutron readings on the data 

sheet. In the other instance, NIOSH reviewed three data sheets, one from November 1959, 

another from January 1962, and the third from April 1959. These three data sheets show a total 

of six neutron readings; 10, 20, 20, 20, 20, and 40 mrem. NIOSH determined the MRL value to 

be 20 mrem. This approach is not thorough and supported. Ten neutron readings in one data 

sheet from March 1958 could not represent reasonably the MRLs from 1951 to 1957. It is 

possible that the MRL was higher before 1959. Similarly, it is not reasonable to use six neutron 

readings to represent all neutron measurements between 1959 and 1976. NIOSH needs to 

evaluate the history of MRL values in those years in more detail. 

In addition, for NTA film dosimeters, NIOSH’s values of 14 mrem and 20 mrem appear low and 

are inconsistent with generic values given for NTA dosimeters, as well as values cited by other 
 

 

 

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DOE facilities with similar neutron source terms. For example, Section 2.2.2.1 of OCAS-IG-001 

provides the following generic MRL value as given in the following statement: 

Watson (1959) discussed the neutron monitoring practices at the Hanford 

facilities and reported the limit of detection/reporting limit for the neutron 

dosimeter in the early 1950s was 90 mrem.
 

For the Rocky Flats Plant, Table 6-17 of ORAUT-TKBS-0011-6 cites the following periodspecific limits of detection. 

Period LOD 

1951–1958 400 mrem 

1959–1960 128 mrem 

1961 120 mrem 

1962–1963 369 mrem 

1964 320 mrem 

1966–1970 120 mrem 

Several major DOE sites, including Hanford and Savannah River, have acknowledged the 

limitations and deficiencies of NTA film, along with the recommendation of using locationspecific neutron-to-photon ratios as a surrogate means for assigning neutron exposure. As for 

the albedo TLDs, NIOSH does not provide any information concerning the detection threshold 

values or MRL used from 1976 to the present time. NIOSH should provide supporting data for 

the MRL value of 15 mrem listed in Table 6B-2. 

Similarly, the TBD does not provide any information concerning uncertainties associated with 

these MRL values. It is very important for NIOSH to provide guidance to the dose 

reconstructors in arriving at a set of claimant favorable MRLs for the missed dose calculations, 

especially when the approach used to determine the MRLs is not very sound. 

5.1.4.2.2 Failure to Properly Address Neutron Exposures at INL Reactor Facilities 

The TBD presumes that neutron exposures at INL’s reactors are “. . . generally not a problem” 

and may, therefore, be ignored. But INL had a total of 52 reactors, most of which were 

experimental/prototypes in design. For example, in the 1950s, INL began testing of 

experimental reactors as part of its Aircraft Nuclear Propulsion (ANP) Program. Between 1953 

and 1961, three reactor assemblies were used for a variety of tests that evaluated reactor control 

systems, various nuclear fuels, and the consequences of potential system failures. The testing 

program for the three HTRE (Heat Transfer Reactor Experiment) assemblies was designated as 

Initial Engine Tests (IETs). There were a total of 26 IETs in all. Due to the very nature of 

unprecedented engineering designs and ANP test objectives, these reactor prototypes were 

designed for high power densities, and with minimum shielding and neutron moderation. 

It is understandable that there are difficulties in assessing neutron spectra for this complex array 

of reactor facilities/operations for the purpose of defining correction factors for NTA film 

dosimeters. However, the current approach for assessing neutron exposures at INL reactor 
 

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facilities is clearly inadequate. NIOSH should revise the missed neutron dose approach in this 

TBD. 

5.1.4.2.3 Neutron Calibration Deficiencies 

Based on information presented in ORAUT-TKBS-0007-6, NTA dosimeters were exclusively 

calibrated by a PoBe source, which generates high-energy neutrons. While this may provide a 

suitable calibration for select dosimeters used to monitor workers at these calibration laboratories, 

for other neutron sources with lower energy spectra, neutron doses must be expected to be 

significantly under-recorded (if recorded at all). In brief, NTA dosimeters must be calibrated to 

a neutron spectrum that closely resembles that of a given workplace. NIOSH needs to evaluate 

the missed neutron dose to workers because of this calibration deficiency. 

5.1.4.2.4 Completeness and Quality of INL Neutron Dosimetry and Record-Keeping Programs 

Similar concerns about the completeness and quality of INL dosimetry exist for neutron 

measurements as for gamma and beta measurements. The opening discussion of 

Section 5.1.4.1.1 and the following two sections on gamma and beta dosimetry apply as well to 

neutron dosimetry. Again, NIOSH should investigate the deficiencies mentioned with respect to 

detection of all three types of radiation. 

5.1.4.2.5 Uncertainty Estimation 

The main difficulty in assigning uncertainties to neutron measurements is the strong energy 

dependence of the neutron detection systems. To estimate the uncertainties for the INL 

dosimetry systems, it is necessary to know how these systems were calibrated. External Dose 

TBD Table 6-2 (page 18) gives the Facility Neutron Correction Factors (FNCF) by facility; 

however, it is not clear from the text if these factors were previously applied to the recorded 

neutron dose, or whether they should be applied now by the dose reconstructor. It is not 

explained in the text how these FNCFs were obtained. If the albedo dosimeter were well 

“calibrated” for each INL workplace, then the uncertainties applied for neutron dosimetry in the 

field (and not in the calibration laboratory) would be on the order of 70%–100%. 

5.1.4.2.6 Neutron Organ Dose 

For the dose reconstructor to be able to assess the organ doses, it is necessary to make a neutron 

spectrum breakdown for each INL facility with neutron fields into five energy groups; 

E <10 keV, 10 < E < 100 keV, 0.1 < E < 2.0 MeV, 2.0< E < 20.0 MeV, and E > 20.0 MeV. It 

seems that the MTR has been well characterized in terms of neutron spectra. However, similar 

information on other reactor projects, including ZPPR and TREAT, is missing. NIOSH should 

provide information and guidance for dose reconstruction for workers at the latter two reactors. 
 

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5.1.4.2.7 High-Risk (Dose) Jobs 

The TBD has indicated that sources of neutron exposure at INL include neutron sources at the 

instrument calibration laboratories and 14-MeV neutron generators used to characterize waste. 

NIOSH believes most of the reactors built at INL had no beam ports, and neutrons were 

generally well contained and kept away from the workplace. NIOSH also believes that neutron 

fields will be attenuated much more quickly in concrete or water than will gamma fields, so 

neutron fields are generally not a problem at an enclosed reactor. However, the TBD identifies 

MTR, APPR, and TREAT as potential problem neutron sources because of their beam ports and 

neutron beams. In fact, other reactor facilities, like ETA and ATR, also had neutron beams. 

NIOSH should treat these facilities as well. 

In addition, NIOSH should develop a list of high-risk (dose) jobs at INL facilities, and provide 

corresponding neutron dose rates and worker job doses. This information will be helpful for 

dose reconstructions. Since there are also many employees working at these facilities who may 

not have been monitored by wearing neutron dosimeters, it is essential for dose reconstruction to 

develop missed dose for unmonitored workers. 

5.1.4.2.8 Multiplying Factors for Missed Neutron Dose 

For calculating missed neutron dose when the MRL for NTA film is taken from Table 6B-2, the 

TBD indicates the neutron dose should be multiplied by 1.25 for most workers or by 2 for 

workers on the MTR experimental floor. NIOSH should provide data to support these two 

factors. 

In addition, the TBD recommends assigning 50 mrem of neutron dose to the few office workers 

without neutron dosimetry at the TAN Warm Shop office for each period they were exposed to 

the fuel storage casks. NIOSH should provide data to support this fixed missed neutron dose 

default value. 

5.2 OTHER OBSERVATIONS 

In addition to the specific findings and observations that appear in other sections of this report, 

there are several general observations that can be made. These can be distilled into the following 

two broad observations: 

Observation 1: The content and organization of the INL Site Profile TBDs are quite uneven, 

where useful default values and specific guidance for dose reconstructors are often buried among 

general discussions about site activities, history, and technical information about dosimetry, 

which may or may not have practical application. In addition, there are an insufficient number of 

INL site-specific TIBs to provide guidance similar to those available for the Savannah River Site. 

There is only one (in the occupational neutron missed dose section) example given in the TBDs 

to help dose reconstructors apply the guidance. In some cases, there is no explanation why a 

default value is recommended and why it is deemed by NIOSH to be claimant favorable. The 

INL dose reconstruction workbook, which NIOSH provided to SC&A after the site profile TBD 

review process began, is just a large spreadsheet containing dose equations and default values in 
 

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various cells. It is not readily comprehensible or user-friendly. There are no help pages to assist 

dose reconstructors in walking through data entry, default values and parameters selection, and 

dose calculation process. We also understand that there is no accompanying manual to the 

spreadsheet. While we know that the spreadsheet is intended for viewing and use only by the 

cognoscenti performing claimant reviews, some guidance and explanation should be provided to 

ensure that it is being used correctly. 

Observation 2: The four authors of the INL TBDs and TIB are the most site-specific 

experienced and knowledgeable technical experts working on the development of the INL site 

profile. These four authors were intimately involved in the development, management, and 

implementation of the INL Radiological and Dosimetry programs in the past and the present. 

They understandably demonstrate confidence in the INL radiological protection programs. They 

relied heavily on their own experience and knowledge, and past INL environmental reports and 

dosimetry records in developing the TBDs, and used many of the reports that they personally 

authored as reference material. They did not interview any present or retired workers before they 

completed their work. They did not review any safety-related audit reports conducted by DOEHQ and DNFSB, but they did interview some of the current INL staff and requested all 

dosimetry records. 
 

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6.0 OVERALL ADEQUACY OF THE INL SITE PROFILE AS A BASIS 

FOR DOSE RECONSTRUCTION 

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

for purposes of evaluating specific issues of adequacy and completeness, and a “horizontal” 

assessment of how the profile satisfies its intended purpose and scope. This section addresses 

the latter objective by evaluating (1) how, and to what extent, the site profile satisfies each of the 

five objectives defined by the Advisory Board for ascertaining adequacy; (2) the usability of the 

site profile for its intended purpose (i.e., to provide a generalized technical resource for the dose 

reconstructor when individual dose records are unavailable); and (3) generic technical or policy 

issues that transcend any single site profile that need to be addressed by the Advisory Board and 

NIOSH. As mentioned in the Introduction, the practice of addressing the same items from 

several different perspectives has led to some redundancy in the report. 

6.1 SATISFYING THE FIVE OBJECTIVES 

The SC&A review procedure, as approved by the Advisory Board, requires that each site profile 

be evaluated against five measures of adequacy; (1) completeness of data sources, (2) technical 

accuracy, (3) adequacy of data, (4) site profile consistency, and (5) regulatory compliance. The 

SC&A review of the INL Site Profile finds that the profile generally satisfies these objectives, 

although several shortcomings and potential issues of varying significance need to be addressed. 

Many of the issues involve a lack of sufficient conservatism in key assumptions or estimation 

approaches, incomplete analyses of data, or incomplete reflection of operational or dosimetric 

history. Key issues are summarized below and in Table 6-1 (which is a duplicate of Table 1-1), 

which provides a matrix representation of the identified issues sorted according to the SC&A 

findings and observations. Detailed evaluation and discussion of these issues is provided 

elsewhere in this report. 

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

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

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

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

within the report that treats each issue. The last column of the table presents three categories of 

potential related regulatory non-compliance concern for the listed issues. These three categories 

are defined again briefly as follows: 

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

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

corresponds to a probability of causation (POC) value in excess of 50%, and assures 

compensability to the claimant. 

• 	 Category 2: The use of upper bound values is limited to those instances where the 

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

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

internal and external exposure pathways have been considered. 
 

 

 

 

 

 

 

 

 

 

 

 

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

where the case cannot be dealt with under one of the previous two categories and a 

detailed analysis is required. 

Table 6-1: Issue Matrix for the INL Site Profile 

Descriptions (a) 

Issue 

Classification 

Objective 1: 

Completeness 

Of Data 

Objective 2: 

Technical 

Accuracy 

Objective 3: 

Adequacy of 

Data 

Objective 4: 

Site Profile 

Consistency 

Objective 5: 

Regulatory 

Compliance (b) 

Issue 1: (5.1.1.1) Routine Airborne 

Releases Finding (5) X X X X Category 2 

Issue 2: (5.1.1.2) Episodic Airborne 

Release Finding (6) X X X X Category 2 

Issue 3: (5.1.1.3) Direct Gamma 

Exposures Finding (7) X X X X Category 2 

Issue 4: (5.1.2.1) Completeness and 

Quality of INL Internal Dosimetry 

Programs Finding (8) X X X X Category 3 

Issue 5: (5.1.2.2) High-Risk Jobs 

(Internal Exposure) Finding (9) X Category 1 

Issue 6: (5.1.2.3) Calibration of 

Internal Dosimetry Analytical and 

Monitoring Equipment 

Observation X X 

Issue 7: (5.1.2.4) Changes of Internal 

Dose Limits Observation X 

Issue 8: (5.1.2.5) High Fired 

Plutonium and Uranium Intakes Finding (10) X Category 1 

Issue 9: (5.1.2.6) Skin and Facial 

Contamination Observation X Category 3 

Issue 10: (5.1.2.7) Breathing Rates Observation X 

Issue 11: (5.1.2.8) Non-Occupational 

Worker Elimination of DU 

Background 

Finding (11) X X X Category 2 

Issue 12: (5.1.2.9) Unmonitored 

Workers Observation X 

Issue 13: (5.1.2.10) Naval Reactor 

Facility Workers Observation X Category 2 

Issue 14: (5.1.2.11) Plutonium 

Monitoring Observation X X X Category 1 

Issue 15: (5.1.3) SL-1 Accident Dose 

Reconstructions Finding (1) X X X 

Issue 16: (5.1.4.1.1) Completeness 

and Quality of INL Beta/Gamma 

Dosimetry and Record Keeping 

Programs 

Finding (8) X X X Category 3 

Issue 17: (5.1.4.1.2) Penetrating and 

Non-Penetrating Doses Finding (4) X X X X Category 3 

Issue 18: (5.1.4.1.3) Correction For 

Beta Doses Observation X X 

Issue 19: (5.1.4.1.4) Angular 

Dependence Correction Factor for 

Gamma Dose 

Observation X 

Issue 20: (5.1.4.1.5) Restating Beta 

Dose As Gamma Dose Observation X 

Issue 21: (5.1.4.1.6) Photon Spectrum 

Split Observation X X 

Issue 22: (5.1.4.1.7) Immersion Dose Observation X X X 

Issue 23: (5.1.4.1.8) High-Risk Jobs 

(Beta/Gamma Exposure) Finding (9) X Category 1 

Issue 24: (5.1.4.1.9) Extremity Dose Observation X Category 2 

Issue 25: (5.1.4.1.10) Discrepancies 

between PIC and Film Reading Observation X 

Issue 26: (5.1.4.1.11) Minimum Observation X X 
 

 

 

 

 

 

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Table 6-1: Issue Matrix for the INL Site Profile 

Descriptions (a) 

Issue 

Classification 

Objective 1: 

Completeness 

Of Data 

Objective 2: 

Technical 

Accuracy 

Objective 3: 

Adequacy of 

Data 

Objective 4: 

Site Profile 

Consistency 

Objective 5: 

Regulatory 

Compliance (b) 

Detection Limit 

Issue 27: (5.1.4.1.12) Minimum 

Reporting Level (Beta/Gamma) Finding (3) X X X Category 2 

Issue 28: (5.1.4.2.1) Minimum 

Reporting Level (Neutron) Finding (3) X X X Category 2 

Issue 29: (5.1.4.2.2) Failure to 

Properly Address Neutron Exposures Finding (2) X X X X Category 1 

Issue 30: (5.1.4.2.3) Neutron 

Calibration Deficiencies Finding (2) X X X Category 3 

Issue 31: (5.1.4.2.4) Completeness 

and Quality of INL Neutron 

Dosimetry and Record Keeping 

Programs 

Finding (8) X X X X Category 3 

Issue 32: (5.1.4.2.5) Uncertainty 

Estimation for Neutron Doses Observation X X X 

Issue 33: (5.1.4.2.6) Neutron Organ 

Dose Observation X 

Issue 34: (5.1.4.2.7) High-Risk Jobs 

(Neutron Exposure) Finding (9) X Category 1 

Issue 35: (5.1.4.2.8) Multiplying 

Factors for Missed Neutron Dose Observation X X X X Category 3 

Table Notes: 

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

(b) Category 1: Least challenged by any deficiencies in available dose/monitoring data are dose reconstructions for 

which even a partial assessment (or minimized dose(s)) corresponds to a probability of causation (POC) value in 

excess of 50%, and assures compensability to the claimant. 

Category 2: The use of upper bound values is limited to those instances where the resultant maximized doses 

yield POC values below 50%, which are not compensated. For this second category, the dose reconstructor 

needs only to ensure that all potential internal and external exposure pathways have been considered. 

Category 3: The most complex and challenging dose reconstruction represents claims where the case cannot be 

dealt with under one of the two other categories. 

6.1.1 Objective 1: Completeness of Data Sources 

The breadth of data sources used as a basis for the INL Site Profile is evident in the 287 

reports, papers, and other documents cited as references, including a number of authoritative 

historical documents dating back to the start of site operations in the early 1950s. Based on a 

review of the INL Site Description TBD (Rohrig 2004s), it is evident that NIOSH effectively 

compiled and characterized activities and operations at 14 areas and 101 facilities and processes. 

In fact, this review cites the breadth of operational data provided in several places as a strength. 

Also noteworthy is the use of the minimum reporting levels as a simple approach (Taulbee 2002) 

for calculating missed external doses for workers that are provided in Attachment 6B of the 

Occupational External Dose TBD (Rohrig 2004e). Notwithstanding the general excellence of 

the data sources, SC&A identified a number of areas to be deficient: 

(1)	 The workers interviewed by SC&A as part of the site expert interview process 

(Attachments 2 and 3) characterized the site as having areas and jobs capable of 

delivering acute doses to workers. SC&A found that the TBDs lack separate 

characterization and in-depth consideration of high-risk/dose (acute) jobs at INL facilities. 

Although extensive descriptions of key operations and processing facilities are included 
 

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in the TBD, high-risk or high-dose jobs in the INL facilities have not been evaluated and 

potential missed doses are not considered. 

(2) SC&A found a lack of characterization of potential missed neutron exposures to workers 

at the INL reactors in the Occupational External Dose TBD, even though INL had a total 

of 52 reactors, most of which were experimental/prototype in design. These reactors 

often were built with high power densities and with minimum shielding and neutron 

moderation. It is inadequate to presume that there are no missed neutron doses at INL 

reactors. 

(3) SC&A found a lack of characterization of contaminated soil or materials stored outdoors 

at INL facilities, such as RWMC or ICPP, in the Occupational Environmental Dose TBD. 

For example, dry contaminated evaporation ponds at the reactor and fuel processing 

facilities are not characterized. The contaminated soils or materials that are present may 

be resuspended by blowing winds and vehicular activities, and inhaled by unmonitored 

employees working outside the facility buildings. In addition, the TBD did not evaluate 

the adequacy of the environmental monitoring instrumentation and validity of its 

collected data. 

(4) SC&A found a lack of characterization of potential worker exposures at the High-Level 

Liquid Waste Tank Farms (at TAN, ICPP, and TRA), and in remediation and waste 

management in general. The list of radionuclides provided for those operations is 

incomplete and increases the potential for missed dose. 

(5) The authors of the TBDs demonstrate their general confidence with the radiological 

protection practices, environmental monitoring programs, internal dosimetry programs, 

external dosimetry programs, analytical laboratory programs, quality assurance programs, 

and above all, the dosimetry record-keeping systems at INL over the entire operating 

history. Perhaps as a consequence, the TBDs do not consider potential missed doses due 

to deficient work practices and missing worker dose records. NIOSH did not request and 

review field data and facility specific records, including field logbooks, RWPs, PEQs, 

incident reports, occurrence reports, and contamination reports that may provide data and 

information for missed worker doses. It is important for NIOSH to validate that all 

pertinent records (such as incident report, personnel contamination reports, over-exposure 

reports, doses received at other DOE facilities due to temporary assignments, and other 

records essential for the dose reconstruction) are included in the worker files provided by 

DOE to the claimants. 

(6) SC&A also found an inadequate characterization of worker internal exposures. The site 

profile evaluated the recorded worker internal doses only for the period between 1992 

and 2000, which does not cover the more problematic early years of the INL site 

operation from 1949 to 1991. Instead, the TBD evaluated urinalysis results for 1959, 

1960, and 1961, and whole-body counting results for 1963. The TBD also provides in 

vitro for urine samples and in-vivo MDAs for the entire site history. These data are not 

sufficient to provide any comparison or support for the missed worker internal doses. 
 

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(7) The TBDs do not characterize or provide any information on the potential missed worker 

doses due to extremity exposure, skin contamination, facial contamination, and ingestion. 

Many of these potential missed worker doses could be found at various facilities, such as 

ICPP and SMC. 

(8) The TBDs also lack characterization of the potential missed internal and external doses 

for the hundreds of rescue and cleanup workers involved with the SL-1 accident that 

occurred in January 1961. It is a fact that the equipment used and the radiological control 

policies in that era were not as advanced and protective as those currently available. The 

TBDs should provide adjustment factors for stay-time used, dose field estimates, internal 

dose results, external dose readings, and contamination level estimates. 

6.1.2 Objective 2: Technical Accuracy 

There are a number of issues identified in the course of this review that may be classified as 

deficiencies in technical accuracy: 

(1) The derivation of the background value of 0.16 μg/L used for subtraction from each 

urinalysis result of uranium prior to assessment of occupational internal dose for SMC 

radiation workers is not technically sound. The baseline background (population) intake 

value was determined by a study of urine samples submitted by non-radiation employees 

working at the SMC facility. A better approach would be to use urine samples from nonINL people in the Idaho Falls area, far removed from any sources of radioactivity. This 

approach would not create a suspected bias due to uranium intake through various 

pathways (inhalation and ingestion) by non-radiation workers while working at the SMC 

facility. During the site expert interviews, the dosimetry staff indicated that they tried to 

use residents from the Idaho Falls area, but no one was willing to sign a liability waiver 

form. In a subsequent study, they used 16 non-radiation workers from the CFA. In 

addition, the selected background value (0.16 μg/L) is significantly higher than the 

national average background value. 

(2) There are deficiencies in the neutron calibration program at INL. Due to the use of a 

PoBe source for neutron calibrations, dosimeters would significantly under measure 

neutron doses from sources with lower energy spectra. These deficiencies could cause 

significant missed neutron doses. 

(3) The technical accuracy of the approach suggested for estimating missed internal doses for 

workers exposed to plutonium or uranium whose intakes (mainly inhalation) were 

assessed mainly by in-vivo counts is not scientifically established. The approach is not 

persuasive in view of the varying age and isotopic composition of plutonium or uranium 

at the INL site (similar to Hanford). 

(4) The overall approach of the Occupational Internal Dose TBD is to use the significant 

radionuclides for ICPP processed fuels as the most limiting (bounding) source terms for 

the calculation of missed worker internal doses. These source terms were suggested to be 

well-tagged with beta-emitting radionuclides, which allowed beta/gamma-detecting 
 

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CAMs to be used at ICPP. As a result, all possible alpha contamination or internal 

exposures would have been detected and monitored. In reality, there are facilities and 

areas at INL, such as the calciners, where alphas sources predominate and beta-emitting 

radionuclides cannot be used as a tag. Hence, the overall approach reported in the TBD 

is not sound and bounding, because it ignores the fact that there were shortcomings in the 

earlier CAM systems and also deficiencies in the internal exposure control work practices 

at ICPP and other INL facilities. Many site experts interviewed indicated that there were 

significant incidents where internal over-exposures and personnel contamination occurred 

at INL facilities that were not monitored or documented. 

(5) The MESODIF model (mesoscale isopleths) used in the Occupational Environmental 

Dose TBD to determine the ground radionuclide concentrations from airborne releases 

due to routine operations or episodic events from INL facilities was found to be deficient 

by the DOE-HQ Tiger Team in 1991 (DOE-HQ 1991). These concentrations were used 

to calculate the default worker intake values in different INL facilities. Therefore, the 

validity of default intake values provided in Tables 4-1 through 4-12 is jeopardized. 

NIOSH should evaluate the deficiencies in the model and determine the associated 

uncertainties. Specifically, the TBD did not evaluate the doses to workers outdoors 

associated with episodic releases to the atmosphere. 

(6) The Occupational Environmental Dose TBD uses the fence-line TLD results provided in 

the environmental monitoring data reports to determine the direct gamma doses from 

airborne releases and their cumulative ground depositions to personnel working outdoors. 

This approach assumes all workers working outdoors at a specific facility would receive 

an average direct gamma dose from normalized ground concentrations. If the assumption 

were valid, the fence-line TLD results should be adjusted by multiplying a weighting 

factor to account for uncertainties in TLD sensitivity and geometry. However, this 

approach is not entirely valid, because it does not take into account the most limiting 

scenarios, i.e., (1) outdoor workers may become immersed in the plume of routine or 

episodic releases from the facility stack; (2) outdoor workers may inhale resuspended 

cumulative ground radionuclide depositions; and (3) the cumulative ground 

concentrations inside the fence line are generally higher than that near the fence line. The 

fence-line TLDs are too far from the bounding source terms to represent the actual direct 

gamma doses received by the outdoor workers. Therefore, this TBD approach is not 

claimant favorable. 

(7) The approach in calculating the missed external doses (both gamma and neutron) by 

using the minimum reporting level equation is not claimant favorable. This suggested 

approach is based on OCAS-IG-001 (Taulbee 2002), which assumes a statistical average 

dose value (MRL/2) for all workers. However, the dose reconstruction for a particular 

worker, especially in a compensable case (with cancer), should use the bounding dose 

value (i.e., MRL instead of MRL/2). In addition, the development of the MRL values 

was not comprehensive in the Occupational External Dose TBD, and NIOSH’s MRL 

values of 14 mrem and 20 mrem appear low and are inconsistent with generic values 

given for NTA dosimeters, as well as values cited by other DOE facilities with similar 

neutron source terms and dosimeters. 
 

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6.1.3 Objective 3: Adequacy of Data 

(1) Questions regarding data adequacy, where they arise in the SC&A evaluation, have 

largely focused on the adequacy of available facility release data and fence-line TLD 

dose values for estimating occupational environmental doses to unmonitored outdoor 

workers, who may have been immersed in release plumes, inhaled resuspended 

contaminated soils, and exposed to direct gamma radiation from accumulated ground 

depositions. 

(2) Since the Occupational Internal Dose TBD does not use dosimetry data, SC&A focused 

on the adequacy of processed fuel characteristic data for estimating missed internal doses 

to unmonitored workers, who more likely have had missed doses due to inconsistent 

monitoring in the field in the early monitoring programs, from high-risk jobs, or 

unplanned intake or contamination incidents. The same is true for missed gamma and 

neutron doses. 

(3) SC&A also found that the TBDs do not fully explore and develop procedures and 

guidelines to dose reconstructors that would lead them to focus on gaps in environmental, 

internal, and external doses that could lead to a significant underestimate of worker dose. 

Input from interviewed site experts indicates that there were situations where reactor 

workers were not provided neutron dosimeters or were not monitored on a continual basis, 

and where processing facility workers were not monitored when they had positive nose 

smears. Data either presented in the TBDs, or on which the TBDs are based, cannot be 

considered adequate unless an evaluation is conducted of the comprehensiveness of the 

neutron-monitoring and bioassay programs, and to what extent existing dose estimation 

assumptions and methodologies address this potential missed dose. 

(4) The lack of actual worker bioassay data during the 1949−1991 period represents an 

important area in which adequacy of data is of concern. There are no data provided in the 

TBD concerning bioassay of high-fired plutonium and uranium. There may be workers 

with potentially high exposures during that period that cannot be adequately 

reconstructed, especially when worst-case dose reconstructions are required. Lack of 

knowledge of uncertainties in the actual bioassay techniques and instruments used to 

quantify internal dose and the minimum detectable activity (MDAs) represents an area of 

data inadequacy that can lead to significant underestimates of worker dose in this period. 

(5) The source term list of radionuclides provided for the current and past environmental 

restoration and waste management projects at INL is incomplete and may contribute to 

missed dose. Risks of exposure to radionuclides that workers have encountered when 

retrieving and processing RWMC’s stored TRU waste have not been adequately 

addressed. 

(6) INL documented worker exposure incidents and unusual occurrences in facility record 

files. Although INL has significant quantities of personnel monitoring data, as well as 

field radiological control data, there are considerable gaps in the information. In many 

cases, some of this information may not be kept in the worker exposure files. In addition, 
 

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there are problems with the adequacy of data, particularly with regard to worker intake, 

high beta exposure, and extremity dose for high-risk (dose) jobs in some facilities, such 

as ICPP and reactors. 

(7) INL maintained worker exposure records by including (some) dosimetry results and 

facility incident reports in the worker files. There are problems with the completeness of 

these worker files, however, with regard to worker intake results, external dose results, 

extremity doses, and contamination reports. 

6.1.4 Objective 4: Consistency Among Site Profiles 

(1) While INL, Hanford, and the Savannah River Site (SRS) had some missions that were 

similar, marked distinctions existed and continue to exist in facility design, operations, 

operational history, and radiological practice. NIOSH has appreciated this distinction 

and tailored its TBD assumptions and analytic approaches to the unique histories and 

conditions at the three sites, while mirroring those assumptions and approaches where 

justified. Both the Hanford and SRS site profiles predate the INL Site Profile; therefore, 

NIOSH benefited greatly from the early efforts at Hanford and SRS, and was able to 

remedy many of the apparent inconsistencies, especially in the SRS TBDs. 

(2) Attachment 4 of this SC&A report provides, in tabular form (Tables A1–A4), an 

evaluation and comparison of the default assumptions for each element of exposure (i.e., 

occupational medical dose, occupational internal dose, occupational external dose, and 

occupational environmental dose) of INL, Hanford, and Savannah River. The lapses in 

consistency noted by SC&A include inconsistent methodologies and assumptions 

regarding external, internal, and environmental dose for almost identical monitoring and 

exposure conditions at the three sites. This comparison shows that the INL TBDs did not 

provide as much characterization of the default internal dose parameters and external 

exposure factors as the Hanford and SRS TBDs. 

(3) An extensive comparison was performed by SC&A to compare and contrast the 

methodologies used in the INL, Hanford, and SRS TBDs to determine external dose. 

This comparison focuses on the methodologies and assumptions associated with dose 

assessments and the derivation of values used to obtain a POC for individual claimants. 

A detailed analysis is provided in Table A-4 of Attachment 4 to this report. This table 

notes where the INL, Hanford, and SRS site profiles differ or agree on a number of 

important assumptions. In summary, where inconsistent approaches or methods exist, 

they typically represent some lapses present in the INL TBD, despite NIOSH’s 

experience with Hanford, SRS, and earlier reviews. 

6.1.5 Objective 5: Regulatory Compliance 

NIOSH has complied with the hierarchy of data required under 42 CFR Part 82 and 

its implementation guides. However, SC&A has identified some significant shortcomings of the 

data used in the review process of the INL Site Profile that may lead to dose reconstructions that 
 

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are not claimant favorable. It is especially crucial for NIOSH to re-evaluate the technical basis 

of the missed dose assumptions used in the TBDs. 

6.2 USABILITY OF SITE PROFILE FOR INTENDED PURPOSE 

SC&A has identified seven criteria that reflect the intent of the Energy Employees Occupational 

Illness Compensation Program Act of 2000, the Final Rule, and the regulatory requirements of 

42 CFR Part 82 for dose reconstruction. Because the purpose of a site profile is to support the 

dose reconstruction process, it is critical that the site profile assumptions, analytic approaches, 

and procedural directions be clear, accurate, complete, and auditable (i.e., sufficiently 

documented). SC&A used the following seven objectives to guide its review of the INL Site 

Profile to determine whether it meets these criteria: 

• 	 Objective 1 − Determine the degree to which procedures support a process that is 

expeditious and timely for dose reconstruction 

• 	 Objective 2 − Determine whether procedures provide adequate guidance to be efficient in 

select instances where a more detailed approach to dose reconstruction would not affect 

the outcome 

• 	 Objective 3 − Assess the extent to which procedures account for all potential exposures 

and ensure that resultant doses are complete and are based on adequate data 

• 	 Objective 4 − Assess procedures for providing a consistent approach to dose 

reconstruction, regardless of claimants’ exposures by time and employment locations 

• 	 Objective 5 − Evaluate procedures with regard to fairness and the extent to which the 

claimant is given the benefit of the doubt when there are unknowns and uncertainties 

concerning radiation exposures 

• 	 Objective 6 − Evaluate procedures for their approach to quantifying the uncertainty 

distribution of annual dose estimates that is consistent with and supports a DOL POC 

estimate at the upper 99% confidence level 

• 	 Objective 7 − Assess the scientific and technical quality of methods and guidance 

contained in procedures to ensure that they reflect the proper balance between 

current/consensus scientific methods and dose reconstruction efficiency 

The following addresses these objectives: 

(1) The INL Site Profile does a very good job in the Occupational Medical Dose TBD in 

assessing the potential organ doses to workers who received medical x-ray examination 

over the entire operating history. The default organ dose values can be easily identified 

and used from Table 3A-1 (Rohrig 2004m). One potential opportunity to improve the 

validity of the dose assessment is to include a multiplying factor accounting for 

uncertainties in equipment parameters. 
 

 

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(2) The INL Occupational Environmental Dose TBD presents a thorough compilation of the 

airborne releases from routine operations and episodic events at INL facilities. The TBD 

also compiles extensive information on the fence-line direct gamma doses for different 

INL facilities. The default worker inhalation intake and direct gamma doses to workers 

can be determined and used from Tables 4-1 through 4-13 (Peterson 2004). The missed 

environmental doses from the airborne releases and fence-line gamma doses should not 

be significant. However, the TBD has not considered several important potential missed 

environmental dose streams, including unmonitored workers inhaling resuspended 

contaminated soils or materials. 

(3) The INL Occupational Internal Dose TBD (Rich and Wenzel 2004) does not mention the 

use of ORAUT-OTIB-0011 (Siebert 2004) for calculating doses from tritium and 

estimating missed doses from this nuclide. There was a small amount of tritium 

produced at INL facilities in early years. Dose reconstructors should be alerted to the use 

of this technical information bulletin (TIB). 

(4) The INL TBD does not address the opportunity to use surrogate (i.e., reference mix) 

radionuclides when data are not available for a less commonly encountered radionuclide 

and thus the 95th percentile cannot be applied in estimating the upper bounds of a like 

dose. 

(5) The INL Internal Dose TBD does not mention the use of an approach recommended for 

other similar DOE facilities when determining maximum dose. This approach is 

provided by NIOSH in ORAUT OTIB-0002 (Rollins 2004). 

(6) The Occupational Internal Dose TBD does not provide bioassay data for the operating 

years before 1992. For the purpose of compiling data needed to reconstruct internal 

doses based on historical operation, NIOSH amassed a considerable amount of data 

describing radionuclides and operations at the various facilities and their associated 

processes. However, NIOSH does not give adequate (or explicit) guidance to dose 

reconstructors on how to navigate through the complex mix of radionuclides required to 

reconstruct historical internal exposures to workers. There are opportunities for 

improvement in the data sets and instructions to the dose reconstructors with respect to 

reconstructing internal exposures. 

(7) The INL Occupational External Dose TBD does not provide external dose data for the 

operating history of the INL site. For the purpose of compiling data needed to 

reconstruct external doses (gamma, beta, and neutron), NIOSH evaluates the MRL values 

of different dosimetry systems (films, TLDs, and neutron dosimeters) used at INL 

facilities. In addition, the TBD provides information on external radiation fields at 

different facilities. However, NIOSH does not give guidance to dose reconstructors on 

how to use these data to reconstruct historical external exposures to workers from highrisk jobs, unplanned over-exposure incidents, and deficient monitoring in the field. There 

are opportunities for improvement in the data sets and instructions to the dose 

reconstructors with respect to reconstructing external exposures. One important 
 

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opportunity for improvement is to include instructions to the dose reconstructors on how 

to treat missed beta exposures. 

(8) The INL Occupational External Dose TBD does not provide guidance to dose 

reconstructors in assessing missed neutron doses at INL reactors. 

6.3 UNRESOLVED POLICY OR GENERIC TECHNICAL ISSUES 

A number of issues were identified that are common to the INL, Hanford, and SRS site profiles 

and, in some cases, represent potential generic policy issues that transcend any individual site 

profile. These issues may involve the interpretation of existing standards, how certain critical 

worker populations should be profiled for historic radiation exposure (e.g., construction workers 

and early workers), and how exposure itself should be analyzed (e.g., treatment of incidents and 

statistical treatment of dose distributions). NIOSH indicates that it may develop separate TIBs in 

order to address some of these generic issues. The following presents those issues identified in 

the INL Site Profile Review that SC&A believes represent transcendent issues that need to be 

considered by NIOSH as unresolved policy or generic technical issues. 

(1) Direction on the applicability and usability of the TBDs and/or TIBs to individual dose 

reconstructions is absent. 

(2) Adequacy and completeness of worker records are essential to claimant-favorable dose 

reconstructions. None of the site profiles address this issue or give direction on resolving 

missing records. 

(3) Site expert testimony indicates that many workers moved from one plant to the next on 

the same site, creating a complicating factor in determining overall exposure. 

Establishment of an accurate worker history is crucial in such cases. This will be 

especially difficult to accomplish in cases of family-member claimants, where the 

survivors cannot be expected to have a good grasp of where the worker was stationed and 

when. 

(4) Statistical techniques used in the application of the data to individual workers should be 

considered. However, using statistical averages may not be claimant favorable, since in 

most compensable cases, they would not provide the upper bound for missed worker 

doses. 

(5) Dose from impurities and/or daughter products in radioactive material received and 

processed at sites should be assessed. 

(6) Assumptions on solubility, breathing rate, and ingestion should be addressed. 

(7) A correction factor for external gamma doses should be considered to account for angular 

dependence of dosimeter sensitivities. 
 

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(8) Direction with respect to consideration of incidents and high-risk (dose) jobs in 

individual dose reconstructions should be provided. 

(9) Availability of monitoring records for subcontractor and/or visitors and potential 

exposure while working on or visiting a facility should be ascertained. 

(10) Dose to construction workers and other early workers should be assessed. 

(11) Unique exposure conditions for decontamination and decommissioning workers should 

be considered. The relative impact of each of these items on dose reconstruction is sitespecific and requires independent evaluation in each TBD. 
 

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7.0 REFERENCES 

AEC (U.S. Atomic Energy Commission), 1969. The Waste Calcining Facility at the Idaho 

Chemical Processing Plant, TD898.2 W32, Idaho Nuclear Corporation (Aerojet), Idaho National 

Engineering and Environmental Laboratory, Idaho Falls, Idaho. 

Allen, D., 2002. Internal Dose Reconstruction Implementation Guidelines, Revision 0, OCASIG-002, NIOSH, Cincinnati, Ohio, August, 2002. 

Becar, N. J., J. F. Kunze, G. D. Pincock, and L. D. Vavleck, 1961. Critical Experiment Tank 

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Hankins, D.E., 1961. A Method of Determining the Intermediate Energy Neutron dose, IDO16655, Idaho Falls, Idaho. 

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Analysis Report EBR-1, Mark I, ANL 6411, February, Idaho Falls, Idaho. 

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Physics, Vol. 9, pp. 177–186. 

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INEEL 2003. Design and Waste Acceptance Criteria for the INEEL CERCLA Disposal Facility, 

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INC 1969. Standard Reactor Operating and Maintenance Procedures for the Spherical Cavity 

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November 14, 2003. 

Kittel J. H., M. Novick, and R. F. Buchanan, 1957. The EBR-I Meltdown - Physical and 

Metallurgical Changes in the Core, ANL-5731, Idaho Falls, Idaho, November. 
 

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LaRue, D. M., and M. R. Dolenc, 1986. Decontamination and Decommissioning Plan for 

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Lofthouse J. H., 1971. Standard Operating and Maintenance Procedure for the Split Table 

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Idaho Falls, Idaho. 

Lewis, L. C., D. C. Barg, C. L. Bendixsen, J. P. Henscheid, D. R. Wenzel, B. L. Denning, 2000. 

Idaho National Engineering and Environmental Laboratory Site Report on the Production and 

Use of Recycled Uranium, INEEL/EXT-2000-00959, September. 

McCaslin, J. W., 1968. Health and Safety Branch Progress Report, Idaho Nuclear Corporation, 

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McVean, R. L., G. S. Brunson, B. C. Cerutti, W. B. Loewenstein, F. W. Thalgott, and G. K. 

Whitham, 1962. EBR-II Dry Critical Experiments, ANL 6462, Idaho Falls, Idaho, February. 

Merwin, S. E., 2005. Assignment of Missed Neutron Doses Based on Dosimeter Records, 

ORAUT-OTIB-0023, Revision 00, Oak Ridge Associated Universities, Oak Ridge, TN, March 7, 

2005. 

Mobley, E. V., 1987. Preliminary Safety Analysis Report for the Decontamination and 

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Decontamination and Decommissioning of the ARVFS NaK, EGG-WM-9150, Revision 1, 

August, Idaho Falls, Idaho. 

NCRP 1996. Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, 

and Ground, NCRP Report Number 123, January, Bethesda, Maryland. 

Novick, M, 1958. Letter to J.R. Horan (Idaho Operations Office), Idaho Falls, Idaho, 1 April. 

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Rich, B. and Wenzel, D., 2004. Technical Basis Document for the Idaho National Engineering 

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Rohrig, N., 2004(i). Technical Basis Document for the Idaho National Engineering and 

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Report Calendar Year 2001, DOE/ID-12082(01), Idaho Falls, Idaho. 

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Levels of External Radiation, ORAUT-OTIB-0007, Revision 00, Oak Ridge Associated 

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Zimbrick, J. D. and P. G. Voillequé (eds.), 1969. 1967 CERT Progress Report, Controlled 

Environmental Radioiodine Tests at the National Reactor Testing Station, Progress Report 

Number Four, IDO-12065, INEEL. 
 

 

 

 

 

 

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ATTACHMENT 1: CONFERENCE CALL WITH NIOSH AND SC&A 

Teleconference of June 29, 2005 

to Discuss SC&A Questions/Comments Regarding INL Site Profile8
 

Introduction 

SC&A sent its initial list of questions concerning the INL Site Profile to NIOSH on June 16, 

2005.9

 On June 29, 2005, the individuals listed below, from SC&A, NIOSH, ORAU, and the 

ORAU subcontractor, Intrepid, discussed the questions during a teleconference. SC&A 

summarized the responses to each question and invited all the participants to review and 

comment on the contents. Some comments were provided and have been incorporated into the 

summary. The questions (designated by “Q”) and responses (designated by “R”) are arranged by 

TBD number following a few general comments. 

Teleconference Participants 

SC&A	 ORAU 

Nicole Briggs Ed Scalsky 

John Hunt 

John Mauro Intrepid (sub. to ORAU) 

Steve Ostrow (moderator) Henry Peterson 

Bryce Rich 

NIOSH Norman Rohrig 

Greg Macievic 

Tom Tomes 

Questions and Responses 

General 

(1) Q: The content and organization of the INL Site Profile TBDs are quite uneven, where, 

often, useful default values and specific guidance for dose reconstructors are buried 

among general discussions about site activities and history and technical 

information about dosimetry, which may or may not have practical application. In 

addition, there are no INL site-specific TIBs to provide guidance, similar to those 

available for the Savannah River Site, for example. What guidance is given to the 

dose reconstructors to use the INL Site Profile TBDs? Are workbooks or other 

summary guidance documents provided to the reconstructors that extract the useful 

information from the TBDs and present it in a concise, accessible form? 

8 The lab has changed names and acronyms several times over the years; INEEL (old) and INL (new) both 

refer to the same national laboratory. 9 Letter from John Mauro (SC&A) to Dr. Lewis Wade (U.S. Dept. of Health and Human Services), June 16, 

2005. 
 

 

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R: The dose reconstructors are not provided with any special instruction per se in the 

use of the TBDs, but they have workbooks available. NIOSH10 thought it had 

provided the INL notebook to SC&A at a meeting a few weeks ago. [note – 

subsequent to the teleconference, SC&A checked and determined that it had 

received the INL notebooks, which are in Excel spreadsheet format.] 

(2) Q: Section 2.1 (pg. 13) of the Site Description TBD notes that the Naval Reactors 

Facility (NRF) is exempted under the EEOICPA. However, it should be 

recognized that workers not connected with the NRF had the potential of 

receiving exposures from the NRF through various pathways (e.g., airborne, 

skyshine, ground-shine), especially since the NRF is centrally located on the site. 

In addition, apparently, some INL personnel also worked at NRF, where they 

could have received an exposure. For example, Section 5.1.3 (pg. 9) of the 

Occupational Internal Dose TBD states: “It is possible that some workers’ 

internal dose could have resulted from their support work at the NRF.” No 

information on missed internal, environmental, or external doses from NRF is 

provided in the TBDs. Nor is information of operational and episodic releases 

from NRF provided. How should a dose reconstructor deal with INL individuals 

exposed from NRF activities whether located outside or inside of the NRF 

boundaries? 

R: Some exchange of personnel between NRF and the rest of the site took place in 

the early days of operation, but the badging system was effective and recorded 

exposures appropriately. NIOSH considered exposure contributions from NRF to 

personnel outside the NRF area and concluded that it was not a significant factor. 

For example, of the 114 site-wide releases listed for 1955, only one, of 310 Ci, 

was attributable to NRF. The Historical Dose Evaluation database lists the NRF 

contribution to overall dose in the “other” category, with less than a 0.4% 

contribution. 

(3) Q: The INL Health and Safety Laboratory published annual reports on environmental 

surveillance, external dosimetry, radiation detection, internal bioassay, quality 

assurance, personnel dosimetry recordkeeping, and research and development. It 

is not clear why: only the 1960 Annual Report is used in the Medical Dose TBD; 

only the 1958 and 1962 Annual Reports are used for the Environmental Dose 

TBD; and only the 1958, 1959, 1960, 1961, 1962, 1963, 1968, 1970, and 1971 

Annual Reports are used for the Internal Dose TBD. No justification is given for 

including some years and excluding others. Are the chosen years supposed to be 

representative? If so, why? 

R: NIOSH looked at all the Annual Reports that were available (they were not issued 

every year) and cited only the ones that provided information cited in the TBDs. 

10 “NIOSH” is used to denote a response from NIOSH, ORAU, or Intrepid. 
 

 

 

 

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Introduction (ORAUT-TKBS-0007-1, Rev. 0, 5/7/04) 

No Questions 

Site Description (ORAUT-TKBS-0007-2, Rev. 1, 7/28/04) 

(1) Q: TBD Section 2.3.1 (pg. 27) on the ICPP Criticality Accident of October 17, 1978 

states that: “The atmospheric Protection System (APS) at INTEC ... significantly 

reduced particulate emissions, and filtered all releases associated with the 

criticality event.” Noble gases, for example, would not have been filtered out; 

how were exposures from such radionuclides determined? 

R: NIOSH looked at the incident report and considered the resulting exposure in the 

Environmental TBD. The evaluation included exposure from noble gases. 

Occupational Medical Dose (ORAUT-TKBS-0007-3, Rev. 0, 5/28/05) 

(1) Q: TBD Section 3.3 (pg. 5) and Attachment 3A (pg. 10) recommend 200 mrad 

(Kathren 2003) as the default value for entrance air kerma for pre-1954. 

(a)	 What is the uncertainty associated with this default value? 

(b) For the pre-1954 period, why is the assumed lateral view air kerma the 

same as the PA view air kerma while the former are more than 40% 

greater than the latter for periods after 1954 (Attachment 3A)? 

(c) Kathren 2003 (Table 3.3-1) recommends 500 mrad as the default lateral 

view entrance air kerma value for the pre-1970 period. Why is a lower 

value of 200 mrad recommended in the TBD? 

(d) In addition, the default values for the other periods in Kathren 2003 are 

significantly higher than the default values recommended in this TBD. For 

example, Kathren recommends 100 mrad for PA view and 250 mrad 

lateral view between 1970 and 1985 while the TBD recommends 52 mrad 

for PA view and 74 mrad for lateral view between 1954 and 1990. What is 

the basis for the lower default values? Which values should a dose 

reconstructor use to obtain claimant favorable results? 

R: (a) As stated in the last line on page 7 of Section 3.5 of the TBD, the 

uncertainty is 30% at one sigma, which is a typical uncertainty value for 

medical x-ray equipment 

(b) This is a “slip-up.” No lateral views were taken in the pre-1954 time 

period. The entry should be replaced by “N/A” 

(c) See the previous answer for the pre-1954 period of the table. For later 

periods, NIOSH reported posterior-anterior and lateral exposures (i.e., 52 

and 74 mrad) based on actual equipment settings and practices and, 
 

 

 

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therefore, did not use Kathren’s default values; (d) See the previous 

answer. 

Occupational Environmental Dose (ORAUT-TKBS-0007-4, Rev. 0, 3/30/04) 

(1) Q: The TBD focuses primarily on inhalation dose from airborne releases; no 

consideration is given to inhalation doses from resuspension, oro-nasal doses, 

ingestion doses, and external doses due to cumulative deposition of these airborne 

releases. How should a dose reconstructor deal with these missed doses? What 

about skyshine from high level gamma sources in adjacent facilities? 

R: NIOSH considered dose from resuspension of radioactive material and 

determined that, in the worst case, the contribution through this pathway is not 

significant; death from asphyxiation would occur before deleterious effects from 

radiation. Oro-nasal and ingestion (transferred from the lungs into the GI tract, not 

encountered by eating contaminated material) pathways are considered in 

determining total exposure. Skyshine contributions to external doses are recorded 

by TLDs. In general, the individual facilities within INL are relatively isolated 

from one another and no site is downwind of another so that activities within one 

facility have little effect on other facilities. For example, the SL-1 accident 

disposal area produces some wind-swept activity, but it is not transported outside 

the facility fence. 

(2) Q: TBD Section 4.1 (pg. 6) states: 

This TBD also addresses direct gamma doses resulting from 

facility operations. In general, these doses, if not controlled by 

management, increase with time and create a facility background 

dose. At INEEL, these facility background doses were recorded by 

film badges infrequently and inconsistently before 1970 and by 

thermoluminescent dosimeters (TLDs) on a routine basis since 

1972. These facility background doses, or facility fence-line doses, 

as they are sometimes called, are a nebulous indication of a dose 

that workers could receive if they inhabited the facility. INEEL 

facility fence-line doses (minus background) are presented for 11 

locations. 

(a) Does “facility background dose” result only from general facility 

operations, or should the facility background dose be the sum of doses 

resulting from facility operations and environmental deposition from 

neighboring facilities? 

(b) Can the fence-line doses represent the facility background doses for 

workers occupying the facility inside the fence-line? 
 

 

 

 

(c) When the background is subtracted from the fence-line doses, is it the 

“natural background” dose that would be present in the absence of INL? If 

so, how and where is it measured? 

R: (a) A TLD cannot distinguish between radiation from the facility and from 

neighboring facilities 

(b) Yes. All operational releases are from 250 ft stacks, so environmental 

doses are relatively uniform throughout the site; (c) Background 

measurements are taken in Idaho Falls, 50 miles from the site. 

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(3) Q: TBD Section 4.1 (pg. 6) states: 

ICPP airborne effluents have been attributed to creating the 

maximum INEEL boundary dose. Considering this fact, it should 

be suspected that ICPP airborne effluent would also be 

responsible for the maximum INEEL worker doses. Calculations 

performed for the INEEL TBD show that although ICPP airborne 

effluent is the most radiologically significant release at INEEL, the 

impact to workers is significantly below the allowable and 

acceptable limit. 

Since this TBD is concerned primarily with doses due to airborne effluents, it is 

logical to assume that if “ICPP airborne effluents have been attributed to creating 

the maximum INL boundary dose,” then those effluents would also be responsible 

for the maximum offsite environmental doses to the surrounding population. 

(a) How could the doses resulting from ICPP airborne effluents be accounted 

for as the maximum INL worker doses? 

(b)	 Is there a possibility that a worker might be working in a facility subject to 

higher airborne effluents from a neighboring facility other than the ICPP? 

(c) Is there any consideration given to episodic releases in a facility other than 

ICPP? 

(d) Shouldn’t the maximum worker doses be the summation of occupational 

environmental doses from airborne effluents, environmental doses from 

facility background (cumulative deposition), direct beta-gamma doses 

during various operational and episodic releases, and internal doses from 

ingestion, inhalation, and oro-nasal breathing? 

R: (a) Evaluations have shown that the ICPP airborne effluents, which contained 

a significant amount of long-lived isotopes, were the greatest contributors 

to the environmental dose 

(b) The facility sites are quite isolated from each other and ICPP was the most 

important contributor to the environmental dose; therefore, it is unlikely 
 

 

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that a worker might be exposed to greater airborne effluents than from 

ICPP 

(c) Yes. All identified episodic releases from all facilities were considered 

and actual meteorological data were used; (d) Yes, they are determined 

this way. 

(4) Q: TBD Section 4.1 (pg. 8) states: 

All of the airborne releases at INEEL that have occurred since the 

beginning of the Site were reviewed and analyzed ... This 

request ... was to evaluate the radiological impact to INEEL 

boundary individuals from airborne releases that had occurred 

since the beginning of operations at the Site. With the help of 

NOAA, which had hourly meteorological data from 1956 to that 

time, analyses were completed for all airborne releases that 

occurred at INEEL. The radiological consequences for an adult, a 

child, and an infant were calculated with Version 4 of the 

Radiological Safety Analysis Computer program RSAC-4 (Wenzel 

1990). The results of the study were published in the Idaho 

National Engineering Laboratory Historical Dose Evaluation 

(DOE 1991[a]); this TBD refers to that report as the INELHDE. 

All releases considered for that report are the basis for the 

releases considered in this TBD. In addition, all the releases 

documented in the INELHDE, operational and episodic, have been 

independently reviewed and found, with minor modifications, to be 

substantially appropriate. 

(a) What are the “minor modifications” to the releases used in this TBD? 

(b) Why are these releases, which were used to determine doses to offsite 

populations outside the site boundary, “substantially appropriate” for 

onsite workers, who are largely inside the site boundary and inside the 

fence-line of a facility? In reality, source terms considered for offsite 

population dose evaluation tend to eliminate short-lived radionuclides. 

These short-lived radionuclides could be significant dose contributors to 

onsite workers. 

R: (a) NIOSH modified releases following examination of SC&A’s evaluation of 

the INELHDE. In addition, following analyses to determine the isotopes 

making the greatest contribution to exposure (releases x ICRP68 dose 

conversion factors), the original set of 56 isotopes was reduced to nine. 

These nine account for over 95% of the total dose 

(b) Some short-lived isotopes were eliminated from the original set of 56 

isotopes since decay reduced their contributions by the time they reached 

ground level following release from high stacks. 
 

 

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(5) Q: TBD Section 4.2 (pg. 10) states: “For worker dose reconstruction, the analyst 

should use default values for the calculation...” The TBD then states in Section 

4.2.1 (pg. 10): “The annual inhaled quantities (Bq/yr) provided in Tables 4-1 

through 4-8 for each of the 8 facility areas are based on known and published 

INEEL annual airborne emissions”; the same section (pg. 11) goes on to say: 

In 1968 and 1969, formal Environmental Monitoring Reports 

(EMRs) report alpha, beta, and I-131 concentrations that can be 

correlated with Table 4-3 (CFA) values. EMRs between 1970 and 

1990 were reviewed for data that could be used for this correlation. 

Results of the comparison, showing that the calculated values of 

Tables 4-1 through 4-8 are in reasonable agreement with 

measured values reported by EMRs, are provided in Table 4-9. 

Figure 4-4 illustrates the variation of the INEEL Environmental 

Monitoring sampling results for the 9-year period 1978 through 

1986. The figure also illustrates the close correlation of 

environmental sample results acquired at distant communities and 

those acquired at the INEEL facilities and the effect of foreign 

nuclear tests and the Chernobyl reactor accident on INEEL 

environmental sampling results. 

(a) How does this correlation support the use of offsite airborne releases to 

calculate the inhalation intake for workers at various INL facilities 

exposed to onsite (facility-specific) airborne radionuclide concentrations? 

(b) How should a dose reconstructor deal with doses from ingestion intake, 

oro-nasal intake, resuspension of ground deposition, and direct radiation 

from ground deposition other than inhalation intake from airborne 

effluents? 

R: (a) As mentioned before, releases were through high stacks so that the 

environmental exposures throughout the INL site were fairly uniform 

(b) As mentioned before, all applicable exposure pathways were considered in 

the dose determination. 

(6) Q: TBD Section 4.2.1 (pg. 10) states: 

Meteorological dispersion factors applicable to each INEEL 

facility were picked from the annual average mesoscale dispersion 

isopleths of ground-level air concentrations as published in the 

environmental monitoring reports published for INEEL as 

described in INELHDE (DOE 1991[a], Appendix B) ... When an 

isopleth for a given year is chosen for a particular facility, such as 

SPERT, that isopleth is assumed to apply to PBF, SPERT-I, 

SPERT-II, etc. If a facility was between two isopleths, the highervalued isopleth was chosen. Yearly isopleth values for each of the 
 

 

 

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eight facilities have been extracted from the annual environmental 

monitoring reports and converted from the normalized annual 

concentration (hr2/m3) to concentrations (Bq/m3) and multiplied by 

2.4E+3 m3/yr (the amount of air breathed occupationally per year) 

to produce activity inhaled per year (Bq) for an occupational 

individual. These are listed in Tables 4-1 through 4-8 for each of 

the facility areas. The annual inhaled quantities (Bq/yr) provided 

in Tables 4-1 through 4-8 for each of 8 facility areas are based on 

known and published INEEL annual airborne emissions. 

(a) When average dispersion factors or average mesoscale dispersion 

isopleths are used, are these calculated ground-level facility annual 

concentrations claimant favorable? 

(b) Do these ground-level facility annual concentrations represent cumulative 

concentrations (i.e., annual deposition plus residual from past years)? 

(c) Are these concentrations fence-line concentrations? 

(d) How are the concentrations used for onsite workers at various locations of 

the facility under consideration? 

(e) What are the uncertainties associated with these average dispersion factors 

and mesoscale isopleths? 

(f) Why are only annual inhaled quantities (Bq/yr) considered? 

(g) How should a dose reconstructor deal with ingestion uptake, direct 

radiation from deposition, oro-nasal contribution, and inhalation intake 

from resuspension of airborne deposition? 

R: (a) The given isopleths are for ground-level concentrations and are 

appropriate to use 

(b) No; they represent annual concentrations and do not include contributions 

from deposition from previous years (found to be negligible). Fallout is a 

more significant dose contributor than what is released from a particular 

facility 

(c) Concentrations are at the referenced facilities. Releases are from 

high stacks, so the whole site is covered relatively uniformly 

(d) Tables 4-1 through 4-8 give isotopic concentrations by year for 

selected facilities 

(e) As stated in the TBD section, the uncertainty is a factor of two for 

operational releases and a factor of three for episodic releases 
 

 

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(f) [question withdrawn] 

(g) previously answered. 

(7) Q: This TBD (Section 4.2.1, pg. 10) uses an annual breathing rate of 2.4 x 103

 m3/yr 

rather than the 2.88 x 103 m3/yr (1.6 m3/hr for 1800 hr/yr) of ICRP 68 or the 8.0 x 

103

 m3/yr (RAC2002, pg. 38) of the NCRP. The TBD assumption appears less 

claimant favorable than the ICRP or NCRP assumptions; please comment. 

R: The chosen breathing rare is applicable to a worker exposed to the environment, 

who is not breathing hard from strenuous labor. NIOSH indicated that this choice 

of breathing rate was made uniformly for consistency for all the evaluated sites. 

(8) Q: Section 4.3 (pg. 22) states: 

External radiation dose at a facility can be created by direct 

radiation from two sources: direct beta/gamma radiation from the 

facility, or gaseous effluents released from the facility or from 

adjacent facilities. In general, direct beta/gamma radiation from 

the facility will increase with time because the general 

contamination of the area will increase. In addition, as a facility 

ages, radioactive sources tend to accumulate at the facility, 

causing the general background to increase with time. 

(a) Were cloudshine and skyshine from adjacent facilities considered as potential 

sources of external radiation dose? 

R: Cloudshine and skyshine were considered wherever applicable; e.g., the 

RWMC. Contributions via these exposure pathways would have been 

recorded by worker TLDs. 

(9) Q: It is not obvious that airborne releases from various facilities of INL would be 

consistent at different locations within the site. The question remains whether 

using the average annual concentrations as the basis for dose reconstruction is a 

claimant-favorable approach or not. Also, can these average annual 

concentrations be used for different individual workers, who worked at different 

times, in different locations, and performed different functions within a specific 

facility? 

R: As discussed, this TBD gives environmental exposures at various facilities on the 

site. If a particular worker stayed primarily in one facility, the corresponding dose 

would be applied. If the worker moved around a lot or it is not known where he 

worked at a particular time, the dose reconstructors would apply the maximum 

dose for the different possible locations. 

(10) Q: Does this TBD take the same approach as the External Dose TBD, i.e., adding 

back the detection limit reading for film badges and TLDs? 
 

 

 

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R: No. The reported values are the difference between readings from the TLDs 

onsite and those in Idaho Falls (background measurement). 

(11) Q: TBD Section 4.3.2 (pg. 24) states: 

INEEL facility air-immersion (beta-gamma) doses could be 

calculated from the noble gas and halogen portions of the 

operational releases, and, if applicable, from the noble gas portion 

of the applicable episodic releases. This calculation should be 

unnecessary because these releases would be recorded in the fence 

line TLD doses presented in Table 4-13. 

The TLDs used at that time were not sensitive enough to record beta doses to the 

skin from noble gas and halogen plumes. What instructions are given to the dose 

reconstructor to compensate for this missed dose? 

R: The dose reconstructor would multiply by a factor of two as instructed in ORAU 

Procedure 6, pages 100–101. 

(12) Q: TBD Section 4.4 (pg. 24), in discussing uncertainty associated with operational 

releases, states: “Discussions with the authors of the INELHDE suggest that 

operational releases, which were monitored, could be low by a factor of not more 

than 2. When the annual normalized ground level concentration values are applied 

to the operational releases, the uncertainty could be increased.” 

(a) Should the dose reconstructor multiply the doses from operational releases 

by 2? 

(b) What is the justification for the factor of 2 multiplication? Is it truly 

claimant favorable? What is the error analysis associated with this factor? 

R: (a) and (b) The dose reconstruction process calculates “mean” doses, then applies 

appropriate uncertainty factors, like the factor of two mentioned in this 

section. The Crystal Ball Monte Carlo program is used to analyze 

uncertainties and a factor of two was found to provide a conservative 

margin. 

(13) Q: TBD Section 4.4 (pg. 24), in discussing uncertainty associated with episodic 

releases, states: “In spite of the original effort to be ‘reasonably conservative’ in 

the exposure estimates, some of the authors, however, have stated that the release 

considered for a particular episodic event might be low by as much as a factor of 

3.” 

(a) Should the dose reconstructor multiply the doses from episodic releases by 

3? 

(b) What is the justification for the factor of 3 multiplication? Is it truly 

claimant favorable? What is the error analysis associated with this factor? 
 

 

 

 

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R: (a) and (b) See the response to question 12 above. 

(14) Q: There are no TIBs providing guidance for the INL dose reconstruction. 

(a) How should a dose reconstructor use Tables 4-1 through 4-13? 

(b) There are only nine radionuclides listed in Table 4-1 through 4-8; how 

should a dose reconstructor compensate for annual intake from other 

radionuclides? 

(c) How should a dose reconstructor calculate ingestion uptake? 

R: NIOSH produces TIBs in response to dose reconstructors requesting assistance; 

the latter has not occurred for the INL dose reconstructions 

(a) Instructions are provided in the TBD. Tables 4-1 through 4-8 provide 

yearly operational intakes by isotope for eight different areas. If a person’s 

location for a particular year is not known, the dose reconstructor should 

select the values for the location giving the maximum exposure. 

Tables 4-

10 and 4-11 provide episodic intakes for several facilities. The 

dose reconstructor would sum intakes for operational and episodic 

exposures 

(b) This has been answered in Question 4(a) 

(c) As mentioned previously, ingestion exposure from eating outside material 

is not significant. However, the dose model does account for material 

inhaled then cleared from the lungs to the digestive system (a negligible 

contribution, though). 

(15) Q: Table 4-13 (pg. 37) presents facility fence direct gamma values by year and 

location as TLD – background. It appears that background is subtracted from the 

total dose to identify the portion attributable to INL operations. It is not clear, 

however, how the background is determined. Is it “natural background” from a 

location far removed from the site or background measured near the site, where 

radiation attributable to site activities would have added to the natural background? 

R: As previously discussed, background is taken in Idaho Falls, far removed from the 

INL site. 

Occupational Internal Dose (ORAUT-TKBS-0007-5, Rev. 0, 10/12/04) 

(1) Q: Consistent with the Savannah River and the Hanford site profile documents, there 

is no consideration given to oro-nasal doses from airborne releases. How should a 

dose reconstructor deal with these missed doses? 

R: This was discussed in the response to Question No. 1 in the Occupational 

Environmental Dose TBD section. 
 

 

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(2) Q: Table 5.1.4-1 (pg. 10) lists radionuclides of concern by year of dose assignment. 

(a) Are these lists reduced from longer lists? If so, what compensations are 

made to account for the missing internal doses? 

(b) In the CEDE column, what do the various CEDE values for each year 

represent? Which CEDE value should a dose reconstructor use for a 

specific radionuclide of concern? 

R: The table is included primarily to provide perspective to the dose reconstructor 

rather than for use in determining claimant exposure. (a) The table includes all 

doses assigned for the nine year time frame for the listed facilities; (b) The CEDE 

values listed by year and facility represent actual recorded internal doses for 

individuals. For example, the six numbers listed for the SMC facility in the year 

2000 represent recorded internal doses for six different people. 

(3) Q: TBD Section 5.2.2 (pg. 12) states that “routine bioassay of radiation workers has 

been conducted since the beginning of the site,” and Section 2.1 of the Site 

Description TBD indicates that the National Reactor Testing Station (NRTS) was 

established in 1949. However, Table 5.2.2-1 (pg. 13) of the Occupational Internal 

Dose TBD begins the bioassay data in 1953. The table does not appear complete. 

(a) What was the bioassay practice between 1949 and 1952, and where is the 

bioassay data presented? 

(b) The reference list does not appear to be complete. Many missing reports 

lead to uncertainties in conclusions. Between 1953 and 1960, the table 

shows annual bioassay requirement; some data, however, indicate biannual bioassay results. From 1973 to 2001, there are no actual bioassay 

data collected or referenced. Why? Where are data from RCIMS, as 

mentioned in TBD Section 5.2.3 (pg. 13)? 

R: (a) Construction began in 1949, but nuclear operations did not begin until 

1951. 1953 data is the earliest that the ORAU team could find 

(b) Table 5.2.2-1 is a reference table summarizing the history of bioassay 

practices at INL by year. Dose reconstructions would rely on actual 

bioassay results in a claimant’s files. In general, bioassays were performed 

depending on a person’s work details. 

(4) Q: TBD Section 5.3.1 (pg. 19) states: “Generally workers were asked to submit to 

bioassay whenever they were in an area in which a CAM alarmed.” 

(a) What are the set points for alarms? 

(b) Could there be missed doses due to high set points? 
 

 

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R: (a) The set points are not known, but, in practice, were set considerably below 

permissible levels. The alarms signaled that a problem may have been 

developing and triggered collection of bioassays. Few of the bioassay 

samples produced detectable uptake levels 

(b) No. See the response to the preceding question. 

(5) Q: TBD Section 5.3.1 on air monitoring states (pg. 20): 

However, in the absence of bioassay data for a known radiation 

worker in one of the functioning facilities as a default, claimant 

favorable, unmonitored dose 100 DAC/MPC-hours for Sr/Y-90 or 

Pu-238 could be assumed per year. This is based upon a consistent 

and standing policy of taking confirmatory bioassay for workers 

exposed to known levels of air activity. However for an MDA for 

chronic activity detection would be <0.01 MPC (in the range of 

10-11 μCi/cc for beta) x 10 ratio of general area to breathing zone 

concentration x 1000 hours of undetected exposure = approx 100 

DAC-hrs or approximately 2 x 105 pCi intake. 

Please comment on our interpretation of this paragraph: It appears to apply to the 

situation where a worker may have been exposed to airborne radiation, but the 

area was unmonitored and there is no bioassay data available for the worker 

(either the worker did not submit a sample, or the record is not now available). 

The approach given, of assuming exposure to a minimum level of detection 

activity, would seem to be valid only if air sampling was conducted, a zero 

reading was obtained, and no bioassay was performed. Absent air sampling and a 

bioassay, any level of activity could have been present and the approach is then 

not claimant favorable. 

(a) What is the basis for 0.01 MPC as the MDA for chronic activity? 

(b) Why is the ratio of general area to breathing zone concentration equal to 

10? Is this ratio for indoor or outdoor activity? 

(c) What is the basis for 1000 hours as the duration of undetected exposure? 

R: (a) SC&A’s interpretation that some areas were not monitored is faulty. INL 

recognized the importance of air sampling and monitored most areas 

accordingly, even the cafeteria. However, NIOSH recognized that the 

quoted paragraph may not be totally clear and will rewrite it when revising 

the TBD 

(b) The 0.01 MPC value is an estimate derived from counting data available from air 

monitoring system filters 

(c) Published reports indicate that the factor of 10 ratio is appropriate; (d) 

1,000 hrs per year is seen as a conservative, claimant-favorable 
 

 

 

 

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assumption since workers move around and do not stay in one place all 

day throughout the year. NIOSH thought that this assumption is also used 

in the other site profiles. 

(6) Q: Table 5.3.2-1 (pg. 21) has sets of columns labeled “statistically significant” and 

“highest results.” Please define and explain these headings. 

R: This table has been taken from 1959–1961 annual reports. These reports do not 

state what statistical assumptions were used, but it is thought that at that time two 

sigma above background was standard. The table indicates that of the thousands 

of bioassays taken each year, only a few yielded statistically significant results. 

(7) Q: TBD Section 5.3.2 (pg. 22), which discusses the change from “no positive 

exposure reported” to “no reportable levels recorded” records, is not clear about 

what was the minimum reportable level for bioassay sample analyses. How 

should a dose reconstructor treat these records during the dose reconstruction 

process? 

R: The section from which the quotation is taken is just for information for the dose 

reconstructor. The minimum reportable levels are provided in subsequent tables. 

(8) Q: Table 5.4-2 (pg. 24) shows internal dosimetry in-vivo MDAs for fecal samples. 

(a) What is the sample size? (mg or cc?) 

(b) The footnote instructs “when sample size is not identified in individual’s 

records, assume the activity is that excreted per day.” Please clarify. 

R: (a) The sample size is not given in the table; the sample analyzed is the total 

sample provided by the individual. In most cases, the claimant’s records 

would contain sample size information 

(b) Assume that the sample referred to is the total amount excreted per day. 

(9) Q: The last paragraph of TBD Section 5.6.1.1 (pg. 29) on the SMC (Specific 

Manufacturing Capability) project to produce DU (depleted uranium) armor for 

tanks discusses natural background uranium excretion, and states that “Urine 

samples submitted by SMC nonradiation worker in 1987, 1994, and 1998 were 

assumed to represent nonoccupational elimination of the SMC worker 

population.” 

(a) Why are only SMC nonradiation workers considered? 

(b) Couldn’t the SMC nonradiation workers have been exposed to uranium 

deposition from airborne releases from SMC and other facilities? 

Shouldn’t a wider population be used to determine the natural background 

uranium uptake? 
 

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(c) Is this type of nonoccupational elimination approach used for other INL 

facilities? 

R: (a) The SMC program was run independently from the other INL programs 

and considered only its own people for establishing background 

(b) NIOSH considered that, but using only the SMC population to determine 

background is consistent with the approach taken elsewhere on the site 

and at other sites (e.g., Fernald) 

(c) No. SMC is the only facility handling depleted uranium. 

(10) Q: Table 5.6.2-1 (pg. 30) indicates that some airborne incidents at various INL 

facilities released noble gases and halogens. 

(a) Was any consideration given to external radiation exposure from these 

releases? 

(b) How should a dose reconstructor estimate these doses? 

R: (a) Yes. Releases though 250 ft (76 m) stacks (primarily Kr) produced doses 

that were recorded as environmental exposures throughout the site 

(b) Guidance is provided in the External Dosimetry TBD. 

(11) Q: TBD Section 5.6.2.1 (pg. 30) on high-enriched fuel storage states that “for 

claimants in which it can be established that they worked in the Building 603 

storage facility for extended time periods, and specific bioassay analyses are 

either not available or insufficient, a claimant favorable default intake of 1000 

DAC-hours per year should be assumed, i.e. 1000 hours/year at 2 x 10-9 μCi/cc.” 

The same request for clarification applies here as for question 5(a) in this section; 

how is this approach valid in the absence of airborne monitoring? 

R: See the response to question 5(a). NIOSH also noted that this section of the TBD 

is being rewritten. 

(12) Q: Section 5.6.4 (pg. 33) discusses the Radioactive Waste Management Complex 

(RWMC). Several major floods over the years and operational practices, 

especially in the earlier days of dumping waste into open trenches and pits, 

resulted in spreading of radioactive contamination across the facility. Not only are 

the workers at the RWMC and adjacent facilities vulnerable to airborne releases, 

they are also subject to potential pathways for ingestion, oro-nasal intake, and 

external/environmental exposure to high level radioactive waste materials. Where 

is guidance on accounting for exposure through these pathways? 

R: This has already been discussed. [see Question 1 of the Environmental Section] 
 

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(13) Q: Section 5.6.4 (pg. 33): As recently as 1995 and 1996, there were reported 

incidents of skyshine at Pit 9 due to disposal of high level gamma-emitting waste 

in Pit 17. This skyshine not only affects the onsite Pit 9 radiation workers, but 

also the offsite non-radiation workers in the Pit 9 Administration Area. Did the 

TBD team evaluate all these exposure pathways and source terms for the RWMC 

facility? 

R: Yes. The exposure pathways are well-documented and were considered. TLDs on 

badged workers would have recorded the exposures and those readings can be 

used to determine exposure to unbadged workers in the same area. 

(14) Q: Tables 5.6.4-1 (pg. 34) and 5.6.4-2 (pg. 35) show the total volume and total 

curies of waste inventory at the RWMC facilities. Have these number been 

compared with 1995 and 1996 RWMC Historical Data Task reports? These 

inventory and concentration values are very significant and could cause 

significant internal doses to workers. How are these values used? 

R: The values appearing in Tables 5.6.4-1 and 5.6.4-2 are straight out of the RWMC 

Internal Dosimetry technical basis document. NIOSH believes that these values 

were taken from the RWMC Safety Analysis. In any event, the values given in the 

TBD tables are primarily to indicate to the dose reconstructor the magnitude of 

the radioactive inventory and would not be used in an actual calculation to 

determine exposure to a claimant. 

(15) Q: TBD Section 5.6.6.1 (pg. 35), on the TRA reactors, states: “The claimant 

favorable recommendation is to use any in-vivo counting data in the claimant files 

directly. If applicable data is absent, the in-vivo MDLs may be assumed, 

consistent with information in Tables 5.1.1-1 and 5.4-3.” 

(a) There is no Table 5.1.1-1. Reference should probably be made to 

Table 5.4-1. 

(b) If actual data is absent, why should MDLs be used? They may not be 

claimant favorable, since, as in episodic releases, exposure levels possibly 

may be above MDL. A preferable approach may be to use surrogate data 

(similar worker in a similar situation); please comment. 

R: (a) Yes; NISOH corrected this discrepancy in the revision of this TBD that is 

currently under development 

(b) NIOSH is in the process of generating a co-worker exposure database for 

all sites that could be used in cases where data are absent. This task will 

eventually result in individual TIBs for the various sites. 

(16) Q: TBD Section 5.6.6.1 (pg. 36) on the TRA reactors, states: “The claimantfavorable position is that there could have been some halogens and particulate 

radionuclides released along with the noble gases.” It continues: “Table 5.6.6.1-1 

contains too many radionuclides for efficient dose reconstruction.” 
 

 

 

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(a) Is I-131 considered the most limiting radionuclides in the list of 

radionuclides of concern in Table 5.6.6.1-1? 

(b) Why is the factor of 1.6 used as the weighting factor to account for the 

total iodine dose? The total activity fraction is 0.834. 

(c) What compensation is provided in the weighting factors to account for the 

rest of the iodine contributions? Are there other radionuclides in addition 

to those in this table? 

R: (a) I-131 produces the greatest dose of the iodine isotopes, which, themselves, 

produce almost 95% of the total dose. Hence, I-131 is a good surrogate for 

the total dose 

(b) An earlier version of the table explained the origin of the factor of 1.6. A 

comparison of iodine doses, using ICRP-68 conversion factors, shows that 

I-131 contributes about 60% of the total dose; the reciprocal of 60% gives 

a multiplication factor of 1.6 

(c) The multiplication factor of 1.6 accounts for the dose contribution from 

the other iodines, which account for about 95% of the total dose. 

(17) Q: TBD Section 5.6.6.1 (pg. 37), on the TRA reactors, states: “The claimantfavorable position is to assign a missed dose based on the MDA for in-vivo 

counting (see Table 5.4-3) for Ag-110m and Ta-182.” Since noble gases are 

involved, should the dose reconstructor take into account the external dose from 

these gases? 

R: Noble gas contributions are recorded by film badges or TLDs. 

(18) Q: TBD Section 5.6.6.2 (pg. 37), on the TRA laboratories, states: “Use of the 

radiologically significant radionuclides for Zr fuel processed at the ICPP in 

Table 5.6.2.5-1 is claimant favorable for evaluating inhalation dose from exposure 

to contamination from the TMI fuel.” How should a dose reconstructor account 

for ingestion and oro-nasal doses? 

R: This has already been discussed. [see Question 1 of the Environmental Section] 

(19) Q: TBD Section 5.7 (pg. 37), on missed dose, states: “If claimant file includes 

positive external dosimeter readings, they should be treated as radiation workers 

and the default internal missed dose is applied as outlined in the table [5.7.1]. If 

no detectable external or internal dose information is recorded, only the 

environmental dose should be included.” What approach should be taken if the 

claimant is a radiation worker, but the claimant’s file does not show positive 

external dosimeter readings? Should the default internal missed dose be applied to 

this worker? 

R: No. If there is no external dose, there also is no internal dose. 
 

 

 

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(20) Q: Table 5.7-1 (pg. 38) presents missed dose default assumptions recommended to 

be used by the dose reconstructors. The recommendations, however, are not very 

clear on how to apply these default assumptions. In the pre-1960 period, the 

assumptions are based on urine gross beta results. Then, between 1961 and 1980, 

they are based on whole body counting results. From 1981 on, they are based on 

bioassay results. There are significant inconsistencies in the foundation of these 

assumptions and approaches. How does a dose reconstructor reconcile these 

inconsistencies involving ingestion only and ingestion plus inhalation/oro-nasal 

results? 

R: Historically, this table has been a problem with respect to clarity. However, dose 

reconstructors have not registered any complaints about its use. The co-worker 

data base process may supersede this table in the future. [See response to question 

15 in this TBD.] 

(21) Q: TBD Section 5.8 (pg. 39), on unmonitored workers, states: “it is recommended 

that workers who have no recorded internal dose and wore a personnel dosimeter 

be treated the same as a worker who was monitored but had no bioassay results 

exceeding reporting levels.” 

(a) This is confusing. Isn’t a worker who wore a personnel dosimeter a 

worker who was monitored? 

(b) In the early days, the reporting levels were much higher due to lower 

sensitivity of the instrumentation. Should compensation be given to this 

worker by adding the reporting level or a portion of it to the total worker 

dose? 

R: (a) “Monitored” in this sentence refers to a worker who underwent a bioassay 

for internal dose. This will be clarified in a future revision 

(b) The co-worker data base process should supersede the TBD guidance. 

[See response to question 15 in this TBD.] 

(22) Q: A July 2003 report prepared by SC&A under contract to the Centers for Disease 

Control and Prevention (Contract No. 200-2002-00367) revealed that the airborne 

emissions associated with several of the Initial Engine Tests (IETs 3, 4, and 10) of 

the Aircraft Nuclear Propulsion (ANP) Program, as estimated by the INEL-HDE 

(Historical Dose Evaluation) task group, were likely to have been underestimated 

as follows: IET 3 – underestimate of total radionuclide release by up to a factor 

of about 3; IET 4 – underestimate of noble gases by up to a factor of about 16, 

halogens by up to a factor of about 7, and solids by a factor of up to about 2; and 

IET 10 – underestimate of total radionuclide releases by up to a factor of about 7. 

It is suggested that occupational environmental doses associated with these 

revised estimates of airborne releases from the ANP Program be taken into 

consideration in dose reconstructions; please comment. 
 

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R: NIOSH looked at the SC&A report supplied in the Advisory Board meeting in 

Boise in August 2004. The report, however, was only a “comment” report and not 

a final report. NIOSH revised the TBD where it felt revision was justified. 

NIOSH would like to see the final report if it has been produced. [It was 

subsequently determined that the report has not been finalized; SC&A will make 

it available to NIOSH when it is released.] 

Occupational External Dosimetry (ORAUT-TKBS-0007-6, Rev. 0, 4/6/04) 

(1) Q: TBD Section 6.3.1 (pg. 8) mentions that “control badges which are used to 

subtract background radiation, have also been located there [at each operational 

areas entrance security gate]. This practice may lead to subtracting environmental 

radiation from site activities reducing the reported doses.” 

(a) How would the dose reconstructor know if environmental doses were 

subtracted from the reported doses, and what he is then supposed to do in 

either case? 

(b) Since the measured “environmental dose” could include contributions 

from air-borne radionuclides after release of fission products, what 

procedure is followed to assure that that the same dose is not ascribed 

twice to a worker? 

R: (a) Environmental doses were subtracted from all reported doses for 

individuals. The dose reconstructor would use Table 6-1 yearly facility 

fence dose values (by year and facility) to determine direct gamma 

exposure by multiplying by (2000 WH/y)/(8760 hr/y). This value would 

be included in the person’s total dose 

(b) [no notes taken] 

(2) Q: TBD Table 6-1 (pg. 9) presents the gamma measurements recorded by the facility 

fence-line dosimeters; this table may be more helpful in the Environmental Dose 

TBD. 

(a) It is not clear what “TLD – Background” in the table’s title means, 

especially as TLD monitoring must have begun in the 1970s. 

(b) The background value for 1952–1972 (not given) is probably equal to the 

average exposures for the years 1973–2002 (123 mrem/year). Explain in a 

table footnote. 

(c) It is not clear if the values stated for 1952–1972 are average values. 

Consider, for example, the SL-1 accident at the ARA II site in 1961. The 

dose at the facility fence for that site in that year must have been much 

higher than the 226 mrem (per year) given in the table. 

R: (a) [no notes taken] 
 

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(b) The “1952–72” row entries represents annual doses at each facility 

(c) Film badges in the SL1 area were moved far away from the reactor site 

after the accident and, therefore, do not reflect contributions from the 

accident in the years after the accident occurred. The dose reconstructor 

would assume 226 mrem/y for each year in that range. 

(3) Q: TBD Figure 6-4 (pg. 12), a dose report, shows an entry of 120 rads from beta 

radiation. Although there was a second, less sensitive, film included in the 

package, this is a very high dose, out of the range of modern film dosimetry 

systems with a maximum net optical density of about 4.5. OCAS IG-001 states 

that the saturation optical density of the Dupont film 502 was around 2.8. The 

TBD should discuss upper detection limits and how the dose reconstructor should 

account for high doses in excess of those limits. 

R: The original document gives 1000 R as the upper range of detection. Therefore, 

the dose of 120 rads is within range. 

(4) Q: TBD Section 6.3.2.1 discusses the film badge system used initially at INL, and 

says: “Type 552 film has a threshold level of about 30 mR, and type 558 film has 

a threshold level of about 10mR” (pg. 13). The minimum detection level of film 

dosimetry depends strongly on the energy of the radiation. 10 mR would be 

possible for low energy photon radiation, but not for energies above about 200 

keV, which are characteristic of what was experienced at INL. 

R: A high-energy radium source was used for calibration. NIOSH maintained that 

published reports indicate that a 10 mR minimum reporting level (MRL) was 

achieved. NIOSH has since supplied supporting documented from the “O-drive” 

to SC&A. 

(5) Q: TBD Section 6.3.2.7 (pg. 17) discusses nuclear track emulsion-Type A (NTA) 

neutron detectors. The second paragraph begins with the statement that “the 

minimum dose assigned was 14 mrem,” then presents anecdotal justification 

based on some 1958 readings. The following paragraph, discussing April 1959 

data, ends with the statement: “These values suggest an MRL of 20 mrem.” It is 

not clear what value the dose reconstructor should assume, or whether assuming 

either 14 or 20 mrem would be claimant favorable. 

R: The 14 mrem value applies through 1958 and the 20 mrem value applies 

thereafter. The protocol for reading NTA detectors changed in the 1958 to 1959 

time frame. SC&A noted that the minimum dose assigned was much higher at 

other DOE sites; for example, it was taken as 50 mrem at the Savannah River Site. 

(6) Q: Table 6-2 (pg. 18) presents INEEL Facility Neutron Correction Factors (FNCF) 

for Hankins dosimeters for several of the site facilities. It is not clear from the 

accompanying text whether the data recorded in the workers’ records were 
 

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already adjusted based on the FNCFs or whether the dose reconstructor should 

make the adjustments. 

R: The adjustments have been made in the recorded data. NIOSH stated that it would 

clarify this point in a future revision of the TBD. 

(7) Q: TBD Section 6.3.2.8 (pg. 18), when discussing the date of changeover from NTA 

to albedo neutron monitoring, states that “dose reconstructors should make the 

claimant-favorable assumption that this transition occurred in October 1976.” 

Why is this assumption claimant favorable? If the albedo monitors recorded 

higher doses than the NTA monitors, then shouldn’t readings from the latter be 

adjusted upward appropriately? 

R: The multiplier factor applied to the NTA is greater than one, so the assumption is 

claimant favorable. 

(8) Q: Table 6-3 (pg. 20) lists estimated (based on judgment) percentage laboratory 

calibration uncertainties for beta/photon dosimeters. How were these uncertainties 

factored into reported dose in a claimant-favorable fashion? 

R: Individual uncertainties are factored into the IREP process of deriving a claimantfavorable dose. 

(9) Q: Figure 6-8 (pg. 21) shows two curves, but they are not labeled. Also, the 

probability density should fall to zero at zero MeV. 

R: The curves are not labeled. Number “1” is probably measured values and number 

“2” calibration values. NIOSH thought that it was permissible that the graph not 

go to zero at zero MeV since there are some low energy neutrons present. 

(10) Q: Table 6-4 (pg. 21) lists sources of anticipated laboratory bias in calibrating 

neutron dosimeters. Some of the rows indicate that “recorded dose of record is 

likely too low.” How is that finding factored into deriving claimant-favorable 

dose numbers? 

R: This finding has not been factored explicitly and separately into determining 

doses. Rather, all uncertainties are folded into the IREP program. 

(11) Q: Table 6-6 (pg. 23 – note that Table 6-6 precedes Table 6-5, which is placed on 

page 25) shows IREP beta and photon energy groups for different INL facilities. 

In all cases, save the last, the split between the lower and higher photon energy 

groups is given as 25% to 75%. This assumption is not as claimant favorable as 

the SRS or Hanford photon spectrum assumption of a 50% to 50% split (see, for 

example, SRS ORAUT-TKBS-0003, Table 5.3.4.1-1). Why was the 25% to 75% 

split chosen for INL? 

R: NIOSH will look into this. [subsequent to the teleconference, ORAU team 

member Intrepid determined that while Savannah River used the 50:50 spectrum 
 

 

 

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mix, Hanford and other sites used the 25:75 mix adopted for INL. Intrepid 

believes the latter mix is appropriate since the unshielded dose fraction in the 

upper energy group (> 250 keV) represents over 95% of the total dose, and that 

after passing through some shielding, an assumption of 75% in the upper group is 

a reasonable approximation of that dose mix.] 

(12) Q: TBD Section 6.3.4.2, on beta radiation, refers to Figure 6-10 (pg. 24) as indicating 

that the beta exposures can only be appropriately applied to the skin and not to the 

breast or testes. The figure, however, shows that 100% of the nuclides will 

contribute to the 0.07 mm dose (skin), but that also at least 30% could contribute 

to the 1 mm (100 mg/cm2

) dose at the closest parts of the breast and testes. 

Therefore the potential effect of beta exposures to the breast and testes should not 

be disregarded. Also, it is not clear whether the general shallow dose procedures 

developed by NIOSH should be applied at INL, namely the “Addendum to 

External Dose Reconstruction Procedure: Shallow Dose Calculations for 

Complex-Wide Cases,” ORAUT-PROC-0006 Rev. No. 00 PC-2. 

R: It is not appropriate to apply beta surface dose to the testes and breast at the 1 

mm level. That there is some beta contribution at that depth is factored into the 

dose reconstruction process. 

(13) Q: Table 6-5 (pg. 25) of the TBD presents data to allow determination of correction 

factors for the recorded beta doses to compensate for the underreporting 

associated with dosimeter thickness. It is not clear, however, how the table values 

were obtained. 

R: The correction factors account for the dosimeter covers and dosimeter thicknesses. 

A dose reconstructor would divide the reported dose by the factor given in the last 

column. 

(14) Q: TBD Section 6.3.4.3.1 (pg. 27) states that: “The exception to the above 

discussion [no beam ports – low potential for neutron exposure] is the MTR, 

which ... had beam ports and neutron beams extending onto a research floor. Also 

in this category are the Zero Power Physics Reactor (ZPPR) and Transient 

Reactor Test (TREAT), both at Argonne West.” Why was the Neutron 

Radiography (NRAD) TRIGA reactor (described in Section 2.4.13 of the INL Site 

Description TBD, ORAUT-TKBS-0007-2) not included as well? 

R: The TRIGA is excluded since it was installed in 1977, after NTA dosimeters were 

no longer used. 

(15) Q: TBD Section 6.5.2 (pg. 33), on neutron missed dose, states: “If no neutron dose 

was assigned to him/her or to co-workers for several months, the dose 

reconstructor should assume that the person was not exposed to neutrons so no 

neutron dose would be missed.” It is not clear if these instructions pertain to all 

workers or only workers in the designated neutron exposure areas (MTP, ZPPR & 

TREAT). For those workers in the neutron exposure areas, it may not be claimant 
 

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favorable to disregard a series of zero measurements since the workers may have 

been exposed to ongoing low levels of radiation below levels of detection. Also, it 

appears that the term “several months” could be interpreted in various ways by 

the dose reconstructors, resulting in different missed neutron dose estimates. 

R: “Several” may be interpreted as more than three. The dose reconstructor examines 

a copy of the dose report, which includes co-worker data along with the 

claimant’s exposure. If they all have zero neutron dose recorded, then the 

claimant probably also should have a zero neutron dose. Since neutrons are only 

present at a few known areas, it would be inappropriate to ascribe a neutron dose 

to the claimant. 

(16) Q: TBD Section 6.7 (pg. 34) discusses uncertainties and ascribes the following 

uncertainties: photons – 35% at 1 sigma; beta – 50% at 1 sigma; neutrons – 60% 

at 1 sigma. It is not clear how these values were obtained or how the reconstructor 

should apply them to ensure claimant-favorable results. 

R: These are “best guess” numbers, not calculated values. Some recent field 

measurements and Hanford results support these assumptions for uncertainties. 

Uncertainties are incorporated into the IREP 
 

 

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ATTACHMENT 2: ADVANCE QUESTIONS FOR SITE EXPERT 

INTERVIEWS 

INL Plant Site Expert Interview 

External Dosimetry 

(1) The MDL for a dosimeter would be affected by the exposure geometry and potential 

shielding of the dosimeter by the body or other aspects of the geometry of exposure 

relative to the badge location. How was angular dependence accounted for in nontraditional exposure situations? Were angular dependence issues considered with neutron 

dosimetry as well? 

(2) Were there any studies done on the differences in badge response between the calibration 

sources and the actual radionuclides in the field? Are these differences in source term 

accounted for in calculation of dose from dosimeters? 

(3) Who was responsible for performing tests on new badge systems? How were these tests 

documented? Can we obtain a copy of these documents? 

(4) How were individuals directed to wear their beta/gamma dosimeter? How were they 

directed to wear their neutron dosimeter? 

(5) What were the procedures for assigning dose due to a lost or damaged dosimeter? 

(6) Where were badges stored historically? Were individuals allowed to take them home? 

(7) How did you measure slow neutron dose prior to the TLND? What was the effectiveness 

and detection limit on this system? 

(8) Were temporary badges ever assigned to individuals on the routine monitoring program? 

If so, under what conditions? How were the results accounted for when assigning 

external dose of record? 

(9) When is/was extremity dosimetry assigned to workers? 

(10) Are/were there conditions where partial body exposures (e.g., skin, lens of eye, gonads, 

chest, etc.) would have significantly exceeded whole body exposure measured by the 

routine dosimeter? If so, how did INL address this issue? 

(11) Under what conditions, do/did INL use multiple dosimetry? How was the whole body 

dose assigned in these circumstances? 

(12) Are/were dosimetry requirements varied for high-risk jobs, such as during sampling of 

process material or maintenance? 
 

 

 

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(13) Was/Is there an area dosimetry program at INL? If so, when and how were these results 

used? 

(14) Was timekeeping ever used as a mechanism to limit internal or external dose? If so, 

under what conditions and where are the records located? 

(15) Were zeros ever entered into badge records when workers did not turn in their badges? If 

so, how common was the practice? Was it confined to some periods of operation? 

(16) What is/was the process for assigning dosimeters to subcontractors and visitors? 

(17) Are you aware historically of unsanctioned practices by workers with respect to their 

dosimeter (e.g., putting them on sources, not wearing them, etc.)? For instance, were 

there times when workers would take off their dosimeters when the maximum limit was 

being reached so that they could continue to work? If yes, was this kind of practice more 

common in some areas than in others? 

(18) Are there audits or assessments (especially historical) of the INL external dosimetry 

program available? If so, can we get a copy of these documents? 

(19) Can you recommend resources (i.e., technical reports, books, films, etc.) that may be 

helpful in understanding the historical operations of the site? 

INL Site Expert Interview 

Radiological Field Operations 

(1) What was the radiological control organizational structure at the inception of the plant? 

How has it changed over time? 

(2) What is the current radiological organization’s structure? 

(3) Were employees frequently transferred between the different processes or did they 

maintain the same job for a majority of their career? 

(4) How much movement was there between INL and other DOE sites? 

(5) How does this organization interact with subcontractors and/or the environmental 

restoration contractor?
 

(6) What has been the primary regulatory driver (e.g., 10 CFR 835) over the period of 

operation at INL?
 

(7) There are a number of areas at the INL plant where the DOD conducted operations. 

What did these operations involve and where were they conducted? How much support 

did INL workers provide to these operations? 
 

 

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(8) Do you provide RCT, radiological engineering, dosimetry, and instrument support for 

subcontractors and construction support? How do you control the radiological practices 

of subcontractors? How is/was this different in the past? 

(9) When were RWP/SWPs used at INL? What types of RWP/SWPs were used (i.e., job 

specific, routine)? Were early work permits focused on job specific coverage or did they 

include routine work also? Did they have sign in sheets? 

(10) If the data were available, is there a mechanism to associate workers on a particular job 

with survey/air sampling data taken for that job? Why or why not? 

(11) How were/are RWP/SWP requirements such as the following determined? 

• PPE Requirements 

• Bioassay Requirements 

• Dosimetry Requirements 

• Not to exceed radiation and contamination levels 

• RCT coverage 

• Entry/Exit requirements 

(12) In the absence of RWP/SWPs, how were the above requirements determined and 

communicated to the workers? 

(13) How have the posting and contamination control limits changed over the period of 

operation of the site? 

(14) Were there always common procedures for posting, contamination control, and 

determining radiation protection requirements? In other words, was the process for 

determining survey frequency, dosimetry requirements, PPE, etc. consistent over time? 

(15) How did you historically determine monitoring for individuals who work all over the site 

(i.e., maintenance)? 

(16) How are/were spread of contamination and other small occurrences documented? 

(17) How were incidents involving personnel exposure documented? Where are these records 

located? What percentages of incidents are classified? 

(18) What are the radionuclides of concern from a field perspective by building? 

(19) When did the site start receiving recycled fuel? 

(20) In terms of ALARA classification, can you name some high-risk jobs? 

(21) Are there areas on site where administrative offices back up to or are adjacent to 

radiological areas? 
 

 

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(22) What portable survey instruments were originally used at INL? What instruments are 

used today? 

(23) Is there field characterization data? If so, where is this data located? How can we obtain 

a copy? 

(24) How are/were portable survey instruments calibrated at INL? How has this changed over 

time? 

(25) What sources (neutron, beta, gamma) were used for calibration over time? 

(26) Are there documents or reports available on the instrumentation and air sampling 

equipment used now and in the past at INL? If so, how can we obtain a copy? 

(27) What type(s) of air sampling equipment have been used in the past? 

(28) When were breathing zone apparatus first used at INL? Is INL involved in DAC-hour 

tracking? If so, when did this begin? 

(29) How was air sampling equipment calibrated historically? How has this changed over 

time? 

(30) What sort of technology shortfalls have your identified with respect to field 

instrumentation and analytical abilities? 

(31) Describe the engineering controls, administrative controls, and PPE that are commonly 

employed in each building handling radioactive material. How was this different in the 

past? 

(32) How extensive is/was the use of Radiation Generating Devices (e.g., radiography sources, 

x-ray diffraction, etc.)? What types of devices were used? In what operations and 

buildings were these devices used? 

(33) Are there field conditions that are particularly challenging to the field radiological control 

staff? If so, what? 

(34) Have there been particle size and/or solubility class studies at INL? 

(35) What is the relationship between INL and the state of Idaho? 

(36) What is the extent of decontamination and decommissioning operations? What D&D 

activities have occurred in the past? 

(37) Can you recommend resources (i.e., technical reports, books, films, etc.) that may be 

helpful in understanding the historical operations of the site? 
 

 

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INL Site Expert Questions 

Environmental Monitoring 

(1) How were these environmental releases documented? Where are these records located? 

Has INL published any internal reports on releases of radioactive material to the 

environment? If so, where can we obtain a copy? 

(2) Do annual environmental reports include both routine and episodic releases? If not, 

please explain. 

(3) Are you aware of historical particulate and gaseous releases from facilities to the 

(4) What is the extent of outdoor contamination at INL in the soil, groundwater, vegetation, 

etc.? 

(5) When did the environmental monitoring program for soil, groundwater, and vegetation 

begin? What did/does the monitoring protocol involve? 

(6) What is the relationship between INL and the state of Idaho? 

(7) What EPA regulated activities are currently occurring at INL? 

(8) In some situations, liquid effluents can be found to deliver a higher dose to the public 

than airborne releases from a facility. Is this the case for occupational workers at the INL 

site? If so, explain. 

(9) Have you ever calculated environmental doses to onsite workers? If so, where is this 

information documented? 

(10) What radionuclides were released from INL facilities and when did these releases occur? 

(11) Were there any significant episodic releases from INL? 

(12) What controls did INL implement over time to reduce environmental emissions from the 

facility? 

(13) Were there any burning operations at INL involving radioactive material? If so, please 

explain. 

(14) Can you recommend resources (i.e., technical reports, books, films, etc.) that may be 

helpful in understanding the historical operations of the site? 
 

 

 

 

 

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INL Plant Site Expert Questions 

Internal Dosimetry 

(1) Historically, what was the basis for determining who was put on a bioassay program? 

(2) Were there any employees that were exposed to radionuclides who were not on an 

adequate bioassay program? How was internal dose to these individuals determined? 

(3) What were the bioassay requirements in areas where recycled uranium was handled? 

Were individuals tested for potential plutonium intakes? 

(4) Historically, who determined the need for special bioassay samples? 

(5) What areas had the greatest number of intakes historically? Why do you think this was 

the case? 

(6) Based upon current/past decision levels related to in-vivo and in-vitro analyses, have you 

estimated a potential missed dose? If so, where is this information documented? 

(7) How effective were early in-vivo counters for the detection of natural, depleted and 

enriched uranium? Were they used for the detection of other radionuclides of concern? 

(8) Describe the interaction between internal dosimetry and the field personnel during an 

incident or occurrence. 

(9) Have there been any particle size studies done at the INL facilities? If so, where are they 

documented? 

(10) What default assumptions do you used when calculating internal dose (i.e., particle size, 

solubility class, date of intake, type of intake, etc.)? 

(11) Are there areas at INL where there is a potential for an intake of high fired plutonium or 

uranium oxide? If so, how does an intake of high fired oxides affect your internal dose 

calculation? 

(12) How extensively did INL Plant handle tritium? Was there potential exposure to tritides? 

(13) Was there ever a time when air concentration data was used to determine internal dose? 

If so, how was the dose determined? 

(14) Have you done any comparisons between air concentration data, in-vivo data and in-vitro 

data? If so, what were the results? 

(15) Have certain isotopes between used as an indicator for the presence of other isotopes? 

For example, a bioassay program may focus on Cs-137 as the central radionuclide; 

however, when Cs-137 results are positive it may trigger a Sr/Y-90 bioassay. 
 

 

 

 

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(16) How did you account for intakes of a given radionuclide prior to the establishment of 

bioassay techniques for that given radionuclide? 

(17) What is your involvement in incident response? 

(18) What are the background levels of uranium and thorium in fecal and urine samples for 

the geographical area? 

(19) What technology shortfalls have been identified in the current and historical bioassay 

programs? 

(20) Are there audits or assessments of the INL internal dosimetry program available? If so, 

can we get a copy of these documents? 

(21) Are thoron and radon an occupational exposure issue in any areas of the facility? If so, 

where? 

(22) How involved were you in the development of the INL site profile? Did NIOSH/ORAU 

interview you? If so, when? 

(23) Can you recommend resources (i.e., technical reports, books, films, etc.) that may be 

helpful in understanding the historical operations of the site? 

INL Site Expert Questions 

Radiological Records 

(1) Are you aware of any code names encountered in the records and their meanings? If so, 

can we get a copy? 

(2) What were the various terms used to refer to shallow and deep dose through time? How 

were these terms defined? Were tritium and neutron dose incorporated into whole body 

doses? 

(3) Do you have a document describing the historical dosimetry records and the abbreviation 

used on these records? For example, a document that may define codes used to 

differentiate between extremity, whole body and multiple dosimeters. If so, can we 

obtain a copy of this information? If we have questions related to a specific record who 

can we direct these questions to? 

(4) How complete are the dosimetry files (early and current) with respect to dosimeter results 

and calibrations, bioassay results, PIC data, time-keeping data, multiple dosimetry results, 

and investigation reports for lost or damage dosimeters? Are these records associated 

with an individuals personnel radiation exposure file, or are some of the records stored 

with field radiological control records? 
 

 

 

 

 

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(5) Who is in charge of maintaining field radiological records? How are these records stored? 

How retrievable is information such as work permits, early radiological surveys, and air 

monitoring records? 

(6) How are incident and personnel contamination reports stored? Is there an incident 

database that summarizes all incidents at INL? If so, does it include classified incidents? 

(7) How are off-normal situations such as a spill of radioactive material that is not likely to 

result in personnel exposure documented? 

(8) How were/are the doses for subcontractor/construction tracked? Are these records stored 

separately from permanent employee records? 

(9) When are radiation exposure reports issued to visitors, workers, subcontractors, etc? Has 

this always been the practice? For example, if a visitor from SRS came to visit INL and 

entered a radiation area, would INL provide a radiation exposure report to SRS on this 

particular employee? 

(10) Has dose of record been modified for any reason over time? 

(11) Did early contractors take monitoring records with them when a new contractor took over? 

For example, at Hanford DuPont took their dosimetry records and other records with 

them when their contract ended. 

(12) Has the site destroyed any dosimetry, bioassay or field radiological records in the past? 

If so, please explain? 

(13) What radiological exposure records are provided to NIOSH for the purpose of dose 

reconstruction? How often have you received follow up requests for additional 

information?
 

(14) Can you recommend resources (i.e., technical reports, books, films, etc.) that may be 

helpful in understanding the historical operations of the site?
 

INL Site Expert Questions 

Medical Exams/Treatments 

(1) What were the medical exam requirements for workers in the present and historically? 

(2) What did the medical exams include? 

(3) How frequently did INL employees have medical exams? 

(4) How frequently were they given x-rays as a part of this exam? 

(5) What type of x-ray equipment (i.e., manufacturer, model) was used historically? 

(6) What type of x-ray equipment (i.e., manufacturer, model) is used currently? 
 

 

 

 

 

 

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(7) Was photofluorography used at any time at the INL? 

(8) Were 4” x 5” films every used for x-rays? 

(9) Did the radiation safety department ever perform a radiation survey or study to determine 

the exposure from medical exam x-rays? If so, what were these results? 

(10) Who performed x-ray inspections on your equipment? Where are these records located? 

(11) Are there other items that you would like to add that would help evaluate medical x-ray 

exposure? 

(12) How many chelation procedures have been performed at INL? 

(13) How many lung lavage procedures have been performed at INL? 

(14) What is/was the criteria for using chelation therapy? 

(15) When chelation therapy was done, where were the records stored? Did this information 

end up in the medical or radiation exposure file? 

(16) Are there other items that you would like to add that would help evaluate internal 

exposure to radionuclides?
 

(17) Is it possible to review a sample medical record from a long term employee and ask some 

questions? 

INL Site Expert Questions 

Operations/Maintenance 

General Questions 

(1) Which plant did you work in and what tasks did you perform. Did these tasks involved 

radioactive material? If so, what type? 

(2) Were employees frequently transferred between the different processes or did they 

maintain the same job for a majority of their career? 

(3) How many hours a day and per week did you routinely work? 

(4) Did you receive radiological worker training? If so, when did this training begin? 

(5) What particular forms of radioactive material did you work with? 

(6) Were you monitored with dosimeters (i.e., film badges or TLDs) when you were working 

on tasks where radiation exposure was likely? 

(7) Where were you instructed to wear your dosimeter? 
 

 

 

 

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(8) Did you store your dosimeter onsite or take it home? 

(9) Were there situations where a portion of your body was more exposed than others? If so, 

where was your dosimeter in relation to this exposure? 

(10) Did you every wear pocket ionization chambers (also called PICs or pencils) at the plant? 

(11) Did you participate in a urine or fecal bioassay program? If you received sampling, how 

often was it done i.e., monthly, quarterly, annually? 

(12) If you had urinalysis or fecal bioassay, when were you told to submit the sample (e.g., 

after an event, Monday morning, etc.)? 

(13) Were you asked to submit blood samples for the purpose of radiation monitoring? If so, 

please explain. 

(14) Did you ever have a whole body or lung count and if so how often? 

(15) Were you aware of the use of co-worker exposure data to assign a dose? If so, what 

problems or advantages do you see in doing this? 

(16) Were your dosimeter and bioassay results made available to you routinely? When did 

this begin? 

(17) What type of PPE was used throughout the operation of the plant in the various 

operations? Did you wear special protective clothing while working on the job? When 

did you start wearing shoe covers? 

(18) How were PPE and special dosimetry requirements communicated to you prior to the use 

of radiation work permits? Were the PPE requirements consistent between different 

departments for the same type of task? 

(19) Were there situations where an individual in the immediate vicinity of the work wore 

PPE yet an individual standing near that individual did not? If so, please explain? 

(20) Where was support personnel (e.g., health physics technicians, foreman, security, etc.) 

located in relation to the immediate production lines, machining or maintenance? 

(21) Were there showers or change rooms at the facility? Did you shower and change your 

work clothing before leaving for home? 

(22) How were respirators assigned? Did an individual use the same respirator more than one 

time? If so, where was the respirator stored in between uses? 

(23) Did you keep your respirator around your neck in production areas even when not in use? 

(24) Did you support army operations onsite? If so, what was your role? What other INL 

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(25) What operations involved hands-on work with radioactive material verses being 

automated? 

(26) Were you allowed to eat, drink or smoke in the immediate vicinity of the radioactive 

material? Were you allowed to have food or beverages at your immediate work location? 

(27) Did you undergo medical exams and if so what did these exams involve? 

(28) Did you receive medical chest x-rays as part of your medical screening while at INL and 

if so how often was that done? Do you remember any details about these x-rays such as 

film size? 

Operations 

(1) Describe the process for shipping and receiving of weapons, weapons components and 

SNM. Who was responsible for verifying inventory on incoming and outgoing shipments? 

(2) What radioactive materials did INL handle? This would include Research and 

Development activities. What quantity of these materials were handled? 

(3) When did INL start to process and/or store recycled uranium? 

(4) What is the range of enrichment of uranium handled historically and currently at INL? 

(5) How was material transported from one facility to another onsite? 

(6) Were you ever involved in environmental restoration operations? If so, please explain. 

(7) Where there machining operations onsite which involved radioactive material? If so, 

please explain. 

(8) What were the high-risk jobs associated with the operation of the INL site? 

Incidents and Accidents 

(1) How are/were incidents documented at the INL Plant? 

(2) Are there incident reports that are classified? If so, what percentage of the reports? 

(3) How are/were minor occurrences such spills, clothing and personnel contamination, and 

area contamination spreads that could result in internal exposure documented? 

(4) Were you ever involved in an accident, contamination spill, fires, or other incidents? If 

so, please explain? 

(5) How frequent were personnel and/or personal effects contamination incidents? 

(6) Are you aware of other major incidents that occurred at INL? 
 

 

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Environmental Releases 

(1) Were there any significant episodic releases from INL? 

(2) Were there radiological releases from INL facilities? If so, explain. When did this occur? 

(3) Were you exposed to contaminants outside the facilities that might have contributed to 

your dose? If so, please explain. 

Other Questions 

(1) What chemical exposures have you received in conjunction with radiation exposure?
 

Were these chemical exposures simultaneous with radiation exposure?
 

(2) Were you ever directed by management to not wear your badge while around radioactive 

material? 

(3) Were there unauthorized practices occurring at the INL plant? If so, what were they? 

(4) Are there other individuals that should be interviewed? 

(5) Are there sources of historical documents you are aware of which may be beneficial to 

our review? 
 

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ATTACHMENT 3: SUMMARY OF SITE EXPERT INTERVIEWS 

CURRENT AND FORMER WORKERS
 

Interviews were conducted with 24 former workers and approximately 55 current Idaho National 

Laboratory (INL) personnel from operations, maintenance, security, laboratory support, 

environmental monitoring, medical, and radiation protection groups. Personnel represent 

experience at the site ranging from 1954 to the present. The interviews were conducted by 

Ms. Kathryn Robertson-DeMers (C.H.P.) and Dr. Desmond Chan from June 27, 2005, through 

July 1, 2005. The purpose of these interviews was to receive first-hand accounts of past 

radiological control and personnel monitoring practices at INL, to better understand how 

operations were conducted, and to hear worker concerns. Interviewees were selected by the 

EEOICPA site coordinator based on guidance provided by SC&A. Retired workers were 

selected based on their knowledge of historical operations with the assistance of the onsite 

worker representatives. Site experts selected represented a reasonable cross-section of 

production areas and job categories. Interviews with current workers were conducted onsite at 

INL. Retired workers were interviewed in the evenings at the local union hall. 

Workers were briefed on the purpose of the interviews, provided with background information 

on the EEOICPA dose reconstruction program and site profiles, and asked to provide their names 

in case there were follow-up questions. Participants were reminded that participation was strictly 

voluntary. Notes generated as a result of the site expert interviews were reviewed by the Idaho 

Department of Energy Classification office to ensure classified or Unclassified Nuclear 

Information (UCNI) was not inadvertently disclosed. 

INL facilities represented by the site experts interviewed included the following: 

• Aircraft Nuclear Propulsion Program (ANP) 

• Argonne National Laboratory-West (ANL-W) 

• Army Reactor Area (ARA) 

• Advanced Test Reactor (ATR) 

• Battelle Energy Alliance, Inc. (BEA) 

• Burn Out One Two (BOOT) 

• Burial Grounds 

• Bechtel BWXT Idaho (BBWI) 

• Central Facilities Area (CFA) 

• Chemical Research and Analysis (CRA) 

• CH2M-WG Idaho (CWI) 

• Engineering Test Reactor (ETR) 

• Experimental Breeder Reactor I (EBR-I) 

• Experimental Breeder Reactor II (EBR-II) 

• Idaho Chemical Processing Plant (ICPP or INTEC) 

• Loss of Fluid Test Facility (LOFT) 

• Materials Test Reactor (MTR) 

• Power Burst Facility (PBF) 

• Radiological and Environmental Services Laboratory (RESL) 
 

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• Radioactive Waste Material Complex (RWMC) 

• System Nuclear Auxiliary Power Transient (SNAP TRAN) 

• Specific Manufacturing Capability Facility (SMC) 

• Special-Power Excursion Reactor Test (SPERT) 

• Test Area North (TAN) 

• Test Reactor Area (TRA) 

Also included were support personnel who worked throughout the site. 

The job categories represented included the following: 

• Chemical Plant Operators 

• EEOICPA Coordinator 

• Electricians 

• Environmental Monitoring Staff 

• External Dosimetry Staff 

• Firefighter 

• Operational and Support Health Physics 

• Heavy Equipment Operator 

• High-Level Waste Operators 

• Industrial Hygiene and Safety Technicians 

• Instrument Technicians 

• Internal Dosimetry Staff 

• Machinist 

• Mechanical Engineer 

• Mechanics 

• Occupational Medicine Physician 

• Pipefitters 

• Radiological Control Technicians 

• Radiological Control Technician Supervisors 

• Radiological Records Staff 

• Reactor Auxiliary Operator 

• Reactor Operator 

• Security Guards 

• Shift Supervisors 

• Transportation Workers 

• Welders 

• X-ray Technician 

• Yardmen (Laborers) 

Individuals interviewed were given the opportunity to review this summary for accuracy and 

completeness. This is an important safeguard against missing key issues or misinterpreting some 

vital piece of information. Although many individuals provided comments to SC&A, not all 

interviewees provided feedback on their interview summary. Some conflicting opinions arose 
 

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related to documented policies for radiation protection versus actual practice. Both views are 

presented in the summary when available. 

The information the workers provided to SC&A has been invaluable in providing us with a 

working knowledge of the site operations and the radiation protection program. All interviews 

have been documented and summarized below. The information provided is not a verbatim 

transcript, but a summary compiled from multiple interviews with many individuals, and with 

additional background information supplied to improve clarity. In addition, responses have been 

grouped under various categories to improve understanding. Editing by SC&A was limited to 

preserve some of the tone and flavor of the interviews, resulting in some redundancies and lessthan-fluid writing in some places. Individuals have provided this information based on their 

personal experiences. It is recognized that these site expert recollections and statements may 

need to be further substantiated before adoption in the Site Profile TBDs. However, they stand 

as critical operational feedback to the dose reconstruction process. These interview notes are 

provided in that context; site expert input is similarly reflected in our discussion and, with the 

preceding qualifications in mind, has contributed to our findings and observations. 

General Site Description 

The INL covers an area of 890 square miles in the southeast corner of the State of Idaho. The 

INL site is approximately 32 miles wide and 39 miles long. The eastern site boundary is about 

32 miles from the city of Idaho Falls. There are currently 530 buildings onsite. Each INL area is 

very much self-contained, with a full complement of utilities and services in each major area. 

There are about 900 workers at the Naval Reactor Facility (NRF) and 6,000 at INL. Currently, 

the INL site consists of INL (laboratory operations including ANL-W), Idaho Cleanup Project 

(ICP), and NRF. Recently, the Department of Energy (DOE) bought the privatized Advanced 

Mixed Waste Project from British Nuclear Fuels, Inc. (BNFL). 

The Atomic Energy Commission (AEC) operations began at INL in 1949. Prior to AEC 

activities, the site was used as a naval proving ground for ordnance testing. The facility has 

undergone a number of name changes. In 1949, the site was named the National Reactor Testing 

Station (NRTS). It was renamed the Idaho National Engineering Laboratory (INEL) in 1978, the 

Idaho National Environmental and Engineering Laboratory (INEEL) in 1993, and the Idaho 

National Laboratory (INL) in 2005. 

The INL site was divided into three distinct entities; INL, NRF, and Argonne National 

Laboratory – West (ANL-W). NRF was always a separate facility run by the military. ANL-W, 

which was formerly managed by the University of Chicago, was transferred under the 

jurisdiction of the Idaho DOE office in February of 2005. ANL-W and NRF had their own 

support staff (e.g., radiation protection, engineering, maintenance). There were times when 

maintenance personnel were lent to ANL-W or NRF as INL employees. NRF (the Navy) 

gradually separated, depending less and less on the NRTS support. ANL-W basically continued 

to use the NRTS/INEEL support, including the services of RESL. 

In the last 18 months, the site has been undergoing transition from a single contractor to multiple 

contractors. The site has been split into the Nuclear Engineering Program, which is managed by 
 

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Battelle Energy Alliance, and the Environmental Restoration Program, which is managed by 

CH2M-WG Idaho (CWI) and Bechtel BWXT Idaho (BBWI). The Nuclear Engineering Program 

includes the laboratory, research and development, the Nuclear Regulatory Commission (NRC) 

operations, National Nuclear Security Agency (NNSA) operations, and a variety of other 

programs. Many of the facilities onsite have had multiple missions during their course of 

operation. 

The current missions at the INL site include, but are not limited to, laboratory operations and 

environmental restoration. This has created a change of program perspective. There is a 

significant effort to consolidate old programs and facilities. For example, over the past 5 years, 

200 buildings have been demolished; 130 of those buildings have been demolished in the last 

year or two. The program is in a continually changing (“flux”) condition. 

Facility Descriptions 

The INL site has had multiple areas consisting of different programs. These areas included 

Experimental Breeder Reactor I (EBR-I), Argonne National Laboratory-West (ANL-W), 

including the Experimental Breeder Reactor II (EBR-II), Army Reactor Area (ARA) (with SL-1), 

Special-Power Excursion Reactor Test (SPERT), Central Facilities Area (CFA), the Idaho 

Chemical Processing Plant (hereafter, ICPP or INTEC, as it is known now), Test Reactor Area 

(TRA), Test Area North (TAN), Radioactive Waste Material Complex (RWMC), and the Naval 

Reactor Facility (NRF). There have been 52 nuclear reactors built onsite. 

Although the NRF is located on the INL site, this facility is under the auspices of the Department 

of Defense (DOD). In addition to the DOD, the Office of Sciences, the National Nuclear Safety 

Administration (NNSA), and the Nuclear Regulatory Commission (NRC) have done work on the 

INL site. This work is outside the scope of the EEOICPA Program. 

Argonne National Laboratory - West 

The EBR-I area was involved with the experimental breeder reactor program from 1954 into the 

1960s. ANL-W, containing the EBR-II and the Zero Power Reactor (ZPR) facilities, was the site 

of commercial grid reactor testing. ARA housed the Stationary Low Power Reactor No. 1 (SL-1 

reactor) during its brief operation. SPERT was dedicated to testing for reactor excursions and 

reactor safety. CFA housed the support services, such as maintenance, transportation, heavy 

equipment, the laboratory, and health physics services. 

Test Area North 

The earliest program at the TAN area was the ANP, which operated from 1956–1961 for the 

purpose of testing reactors to power jet engines. There were two reactor engines, two rail cars, 

and four rails. The reactors used nickel chromium and ceramic fuel elements. The first reactor 

engine, P102, contained a horizontal reactor with zirconium moderator. The second engine, the 

Core Test Facility (CTF), contained a vertical reactor with a water moderator, and was situated 

on a railcar. The reactors engines were direct cycle, with air circulating through their 

compressors. When an engine was blocked off, the air passed straight through, was compressed 
 

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through the reactor, and traveled back to the other torus. The air then passed through the turbine 

and out a 110-ft stack. Delay tanks and carbon filters were integrated to reduce releases from the 

stack. Chemical fuel was used to turn on the reactor, but once the reactor was going, it provided 

its own power. Since there was no containment around the reactor, essentially, this operation 

became a shielding study project. 

For these two reactor testing facilities, the operator control room was built underground and 

located about 1.1 miles away from the engines. There was a tunnel, free of contamination, 

leading to the engine from the control room and a 110-ton shield built underground for an 

emergency bunker. The engine locomotive could drop down to the shielded bunker in case of 

accident. 

The facility also performed intensive fuel meltdown testing. For example, tests were performed 

to determine the type of emission that would result from burn-up of nickel chromium and 

ceramic fuel. Monitoring stations were established out to 25 miles in all directions from the test 

and many detectors were placed around the site. Some radioactivity was released from the stack 

and deposited in the immediate area. Emissions were much higher with ceramic fuel elements 

than with nickel chromium fuel elements. 

In the ANP Program P102 start-up, the operators overheated (temperature-wise) the reactor and 

the fuel elements became warped. Individuals had to enter the area and manually straighten the 

elements. 

A huge (hanger-size) Hot Shop was built for maintenance and change out of the reactors; the 

largest hanger ever built for such tests. A test dolly built for equipment service purposes was 

used for maintenance, refueling, and servicing the jet engine. The hanger contained a large door 

for the dolly to go in and out. Although the contamination associated with the project was low, 

there were airborne releases from the units. When the P102 was moved to EBR, it experienced a 

mercury leak. Following an accident in January 1961, the SL-1 reactor core was transported to 

the Hot Shop in the TAN area for examination. The remote hot cell facilities in the Hot Shop 

were built for this type of work. 

In 1961, President Kennedy cancelled the ANP Project, along with the Experimental Organic 

Cooled Reactor Project before it went to operation. 

The Loss of Fluid Test (LOFT) facility was located in the TAN area of the site. Personnel 

encountered Sr-90 and Cs-137 contamination during breach-of-system maintenance operations, 

such as cutting pipes. One of the most recent uptakes of Sr-90 occurred last summer during 

Decontamination and Decommissioning operations. An additional hazard at LOFT was mercury. 

The Burn Out One Two (BOOT) test was conducted to determine what would happen if fuel 

burned and completely melted down. The test was to be performed during zero wind conditions; 

however, airborne contamination drifted up the valley and then came back to the site due to the 

wind. The worst emission from this area was as a result of this test. 
 

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In the 1980s, the Specific Manufacturing Capability (SMC) facility was built in the TAN area. 

Operations at the SMC facility, which involved the machining of depleted uranium (DU), are 

considered classified. Chip fires occurred due to the pyrophoric nature of the uranium. An 

incident occurred when the elevated temperature resulting from a worker drilling a can 

containing DU reacted with the moisture inside the can to produce hydrogen gas. The hydrogen 

was ignited by the sparks, and flames shot up to the ceiling. This worker got burned all over his 

face and hands. 

The Initial Engine Test (IET) Facility and the System Nuclear Auxiliary Power Transient (SNAP 

TRAN) project were also located in the TAN area. 

Idaho Chemical Processing Plant (ICPP) 

The ICCP, which began operations in 1953, was built for chemical processing of uranium from 

onsite and offsite fuel. This facility was originally designed as a pilot plant, but was put into 

routine operations. Nuclear fuel rods were dissolved in acid, which produced uranyl nitrates, 

fission products, and some transuranics. Solvent extraction was used to separate the uranium 

from other products. The uranium nitrate hexahydrate (UNH) underwent the denitrification 

process and was converted to powder, producing a fairly pure product. Samples were taken and 

analyzed through the process. Final sampling took place in Z-cell prior to shipment. Initially, 

UNH was shipped to Oak Ridge, and orange oxide was the eventual product of the process. 

The ICPP was composed of several buildings with different functions: 

Building Operation 

601 Fuel and waste processing or reprocessing 

602 Laboratory Services 

603 Old fuel storage building 

604 Rare gas collection plant 

627 Remote Analysis Laboratory 

633 Old Waste Calciner Facility 

637 Laboratory Services 

640 ROVER Dissolution Project 

659 New Waste Calciner Facility 

661 Security area 

666 (FAST) Fluorinel Dissolution Process and Fuel Storage 

The operations were divided into what is referred to as cells or wings. The cells ranged from 

A-Cell to Z-Cell, with the latter the final stage used for packaging of material for shipment. The 

radionuclide concentrations in each cell and the subsequent waste streams were dependent on the 

particular operations conducted in each stage. For example, A-, B-, and C-cells are associated 

with the presence of plutonium. The denitrification cells were associated with UNH and 

yellowcake; at times the yellowcake was visible in these cells. There have been two waste 

calcining facilities built for the processing of high-level liquid wastes. The New Waste 

Calcining Facility (NWCF) was built in the 1970s to replace the older Waste Calcining Facility. 
 

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A process (known as RALA) retrieved radioactive lanthanum from the fuel. This process 

required freshly irradiated fuel that had not been allowed to decay. As a result, there were iodine 

releases associated with this operation. The workers remembered that the temperature of some 

shipping casks coming over from MTR for the RALA Project was extremely high, boiling the 

water in the casks. The casks were also radiologically hot, since the radioactivity in the fuels 

from the MTR reactor hadn’t decayed for any length of time. After the RALA Project, better 

insulated shipping casks were used to move radioactive fuels from Building 603 to Building 601. 

Various other processes occurred at ICPP—neptunium and plutonium were recovered; Building 

601 dealt with recycled uranium, which was dissolved in nitric acid; Navy fuel was processed; 

and the ROVER project was designed to process graphite fuel, where the fuel was burned and 

then dissolved in acid. The project involved multiple curies of U-233/234. 

Raffinates were produced in the process of uranium extraction, and contained fission products as 

well as transuranics. These raffinates were transferred from Building 601 to the liquid tank farm 

for storage through a set of long pipes. Krypton and xenon gases were stripped in Building 604. 

The products were treated with nitrogen to separate out H2, O2, krypton, xenon, and argon. Then, 

H2, O2, potassium, and xenon were boiled off at different temperature ranges. INL had two 

calciner facilities to convert tank waste to crystalline power. 

Test Reactor Area (TRA) 

TRA began operations in the early 1950s for the purpose of material testing. The TRA was the 

location of the Engineering Test Reactor (ETR), the Advanced Test Reactor (ATR), and 

Materials Test Reactor (MTR). The Power Burst Facility (PBF) was an experimental reactor 

used to determine the consequences of reactor accidents; for example, fuel rod failure. 

In general, the facilities in TRA were contained unless systems were breached. The highest 

external exposures seen in the reactor areas are among individuals who performed reactor top 

operations, canal operations, and reactor chemistry. The ventilation at this facility is well 

maintained. Emissions from the stack are filtered and monitored. 

Material Test Reactor (MTR) 

The MTR was the first reactor that went into operation. It was unique in that it has a neutron 

beam port for experimental work. MTR had very heavy concrete walls, due to concerns of 

possible air raid bombing. A huge graphite reflector outside the coolant tank was air-cooled. 

There was a beryllium reflector inside the tank with empty tubes that ran up into the reactor. 

There was also an aluminum reflector that held samples. All the reactors in that era were 

designed to handle the fuels on top of the reactors. 

The MTR top assembly was shielded only by the water tank. To refuel or change fuel elements, 

workers had to pull the top assembly off. Fuel was pulled out of the core and moved under water; 

then the workers would drop it into the discharge chute and it would fall into the canal in the 

basement of the reactor building. During the time the fuel was in the chute, it was not in water. 

In all other operations, the fuel elements were kept under water. The workers used long hooks to 
 

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move the fuel elements. Test fuel immediately removed from the reactor has a dose rate of 

approximately 106 R/hour, followed by rapid decay of short-lived fission products. 

MTR had several horizontal neutron beam ports for experimental work, which could be plugged 

or unplugged to project beams to the ground floor area. The beams were controlled by 

temporary shielding, but they were detectable out to Highway 20. In one case, radiation leakage 

exposed a truck full of film, ruining the film. One beam had what was referred to as a neutron 

chopper. Two disks were tied together with two slots. The chopper speed could be adjusted to 

obtain a particular neutron energy. This unit was used to produce a neutron beam that could be 

used for analysis and research. There were some reported neutron leaks in areas of the reactor 

shield that were not intentional. During one occasion, leakage radiation set off the alarm and the 

area had to be evacuated. 

Experimental Test Reactor (ETR) 

The start-up of the ETR followed MTR. ETR was built to generate from 80 MW to 175 MW of 

power. Used fuel elements were handled under water. Initially, the reactor used flat plate fuel 

assemblies; however, these would ripple and collapse. As a result, the fuel design had to be 

changed. Fuel handling was done underwater. ETR supported the ANP program. 

Neutron flux wires were put in the core and removed for analysis. These were typically removed 

by the reactor operators and/or analytical support. These wires would occasionally break and 

lodge in the heat exchanger. The heat exchanger room was only accessible during reactor 

shutdown. When broken wires lodged in the heat exchanger, this area had high dose rates. 

Eventually, there were heat exchanger problems that required repair, and workers entering the 

area to perform maintenance on the system were exposed to high dose rates. 

Waste material was often stored in the canal. This waste had to be pulled up out of the canal 

water and placed in a cask. This operation required notification of personnel in the office areas, 

and that personnel not involved in the operation evacuate to the north end of the building to 

minimize their radiation exposure. The radiation alarms would sound as the waste was removed. 

The operations took a few minutes to complete. Rarely, a used experimental loop of stainless 

steel was pulled out of the top of the reactor by crane and lowered into the canal. In one incident, 

the loop was dropped and operators involved in the operation were over-exposed. 

Advanced Test Reactor (ATR) 

The ATR, for testing nuclear fuels, was the last reactor built and is the only one currently in 

operation. The ATR is a 250 MW light-water-cooled reactor with beryllium reflectors and 

48-in long fuel at 93% enrichment. The fuel is handled from the top of the reactor. One of the 

missions of the reactor was to produce radioactive materials for commercial use. Capsules 

containing materials were inserted in the beryllium blocks of the reactor along the outer edge. 

Some of the radionuclides produced in these capsules included iridium and cobalt. There were 

no beam ports or neutron leaks at ATR. Hot cells were built for fuel element examinations and 

other testing. Operations of this type continue today. 
 

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There were two sets of operations personnel at ATR. The first was composed of non-exempt 

reactor operators responsible for operations in the canal area of the reactor, including moving 

fuel. Some operators were also responsible for routine maintenance, inspections, and cleanup. 

Exempt operators had jurisdiction over the tanks and the reactor proper, and were responsible for 

keeping the reactor up and operating, and inserting capsules into the reactor. 

Primary system inspections and maintenance were conducted when the reactor was shut down. 

The time post-shutdown for re-entry into areas off limits during operations has varied. Initially, 

personnel were allowed to enter these areas a few hours after shutdown. The loops and primary 

system were not always depressurized prior to entry into these areas, resulting in higher dose 

rates. In the last 10–15 years, re-entry has not been allowed for 3-7 hours after shutdown and 

depressurization of the primary system and loops. During the same time period, an argon delay 

system was installed. 

There were several safety hazards associated with the operation and maintenance of the ATR. 

Exclusion areas at the reactor include the heat exchanger area, the pit area, the area around the 

outer shim cylinder, the rod access area, the pipe corridor, and the Subpile room. The Subpile 

room was probably the most contaminated area at ATR, with contamination the result of 

byproducts that came from the reactor system and, sometimes the fuel. Leaks in pipes, valves 

and packing, and capsule ruptures also resulted in some contamination. Pressurized steam 

represented a substantial safety concern. In general, ATR operations and practices had the 

benefit of learning from mistakes observed at both ETR and MTR. 

Radioactive Waste Material Complex (RWMC) 

The RWMC represents one of the current active areas onsite. The RWMC was designed to 

dispose of waste, both in underground burial sites and in aboveground storage areas. Much of 

this waste was shipped to INL from the Rocky Flats Plant and Mound in barrels or burial boxes, 

and contained considerable amounts of plutonium in a variety of forms. At the burial site, the 

waste containers were dumped into what is referred to as pits or trenches and run over with 

heavy equipment. Site experts observed lids coming off some units, and causing surface and 

in-depth contamination. The primary radiological concern was contamination and airborne 

resuspension of particles. 

There is an ongoing effort to retrieve these compromised barrels, repackage them, and ship them 

to an appropriate facility. Experience thus far has indicated that waste drums and boxes include 

plutonium chips. 

Workforce Characteristics 

The worker population in eastern Idaho was a closed population. There has not been a large 

movement in the last 30 or 40 years except during site management changes. In fact, workers 

are moving to the site rather than leaving the site. For example, workers from Rocky Flats have 

moved to Idaho to work at the INL site. 
 

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The mobility of the workforce onsite was dependent on the period of operation at INL and the 

contractor. At times there was a central maintenance pool located in the CFA area that provided 

support to the entire site. During other periods of operations, support staff was assigned to 

particular areas. Some maintenance personnel worked in teams and were assigned routinely to a 

particular facility. Larger facilities had dedicated maintenance support. With the current 

contractor, maintenance has been reorganized into a central group at CFA. Some support staff 

worked all over the INL site, including ANL-W and NRF. The non-exempt reactor workforce 

was more stable, because of the large investment in training. There was a higher turnover rate 

among exempt staff. 

When there were particularly high dose rate jobs that required limited stay times, supervisors, 

security, and other work groups were utilized to perform simple tasks to prevent maintenance 

personnel from exceeding their allotted maximum dose; often the chosen personnel had no 

training for the job they were told to do. Maintenance personnel were saved for more difficult 

and intricate tasks. For example, in the 1970s, there was a secretary in the ICPP office sent by 

the management to perform a job in the Calciner facility. They gave her a heavy saw and told 

her to cut out a pipe for a valve replacement job. She had no idea what to do or how to do it. 

Other untrained workers would be sent in the hot cells without any training on how to follow hot 

cell entry requirements. Although there was RCT coverage, the RCT was responsible for 

multiple jobs during the day. These management decisions were made by the Operations 

Manager or Production Manager. These untrained employees in the Calciner facility and other 

high dose rate facilities could receive high doses in the range of 3,500 to 5,000 mrem per year. 

The amount of overtime worked depended on the job function and the period of time. 

Historically, there were more overtime hours for maintenance, operations, and security. Part of 

this was due to the limited number of staff available for certain operations. Maintenance and 

operations personnel approximated the number of hours worked per week to be 40–45 hours. 

There were rules established that restricted operations to a maximum of 72 hours per week, 

unless there was a very special situation. Overtime was especially prevalent at ICPP when the 

calciner was in full operation in the late 1980s. The security guards worked up to 70-80 hours 

per week during the 1980s, due to limited staffing. Overtime is limited at the present. 

Security 

Security had a number of responsibilities, including guarding incoming material (e.g., material 

delivered to RWMC), escorting individuals, Special Nuclear Material (SNM) monitoring, facility 

inspections, escort of classified documents, and other routine security tasks, such as traffic 

control. These job responsibilities took security personnel all over the site. 

Material receipt and guarding could be required for up to an entire shift prior to transfer to the 

appropriate facility. Shipments were received from all over the DOE complex and, in some 

cases, from the Navy and commercial entities. There were even foreign fuels shipped to the site. 

Escorting individuals often resulted in questionable exposure conditions. For example, the 

guards were stationed just across the radiological rope from the 603 Cave with no air space 

between them and the work area. The cave door was left open and, while individuals inside were 
 

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in airline respirators, guards outside were simply in their uniforms. Radiation postings were 

variable in some of the areas that guards were required to tour. Tours of the ICPP building 

involved visits every 30 minutes to the denitrification area, all corridors, the sampling areas, and 

analytical laboratories. 

Tackle testing exercises resulted in access to areas where others did not routinely go, such as 

areas at the NRF facility and on facility roof tops. Guards noted that the original area outside the 

fence that was used for some exercises was unposted in the past, but is now posted. As this area 

was not anticipated to cause radiation exposure, many did not wear their film badge, as it was 

easily lost in maneuvers. 

Several facilities in the DOE Complex, such as the Rocky Flats Plant, Brookhaven National 

Laboratory, and the Y-12 Plant, were periodically short-staffed. As a result, INL security 

personnel were sent to these sites to augment the existing staff for short periods of time. In 

addition, security personnel also participated in exercises at Lawrence Livermore National 

Laboratory. 
 

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Radiological Operations 

Organization 

The INL radiological control (RadCon) program was brought in from the Oak Ridge National 

Laboratory (ORNL). The air sampling program, the personal dosimetry program, radiation 

protection program, and some professionals were transferred from the ORNL programs. 

Before 1997, there were multiple companies at the INL site. Former contractors at the site 

included WINCO, Allied Chemical, Exxon, Phillips Petroleum, Westinghouse, Lockheed Martin, 

EG&G, BBWI, and CWI. In the earlier years, Phillips Petroleum was the operating contractor of 

the main site. EG&G later took over this contract, except for the Idaho Chemical Processing 

Plant (ICPP), which Allied Chemical took over. Historically, the site could be divided into four 

entities—ANL-W, ICPP, INEL, and NRF—with a different contractor managing each different 

area. In October 1994, ANL-W was absorbed into INL proper, and in February 1, 2005, it was 

put under the jurisdiction of DOE-Idaho rather than DOE-Chicago. There are also several 

privatized facilities, like the Advanced Mixed Waste Project and the Pit 9 Project. Firemen and 

guards originally worked for the AEC. 

Each contractor onsite was responsible for maintaining a Radiation Safety Program and was 

required to comply with the same regulations. Programs such as dosimetry, instrument 

calibration, radiobioassay services, and environmental monitoring were originally the 

responsibility of DOE-Idaho, but were eventually turned over to the contractor. There have been 

periods of time (i.e., during the Phillips Petroleum through the EG&G and CWI eras) when field 

RadCon personnel were matrixed out to facilities. Historically, individuals working on the 

backshift have reported to the Shift Operations Manager. Fundamentally, the approach to 

radiation protection should have been the same throughout all facilities; however, there were 

facility-level differences. For example, some facilities allowed individuals to wear street clothes 

under their PPEs, while others required company clothing. Some RadCon procedures were 

specific to a certain facility. Currently, field RadCon personnel report to operations at the 

particular facility, and are matrixed to the RadCon organization. 

There is currently only one radiological protection program (RPP) at INL, applicable to 

subcontractors as well. The radiological control (RadCon) organization is currently composed of 

a program director, field RadCon managers, supervisors, Radiological Control Technicians 

(RCTs), radiological engineers, and a central technical support group. RadCon field operations 

support a variety of projects, including nuclear naval program testing, reactor outages, 

maintenance activities, decontamination and decommissioning, and routine area entry and egress. 

The RCTs are responsible for the hands-on radiological protection work, including routine and 

job-specific surveys, job coverage, ensuring workers comply with radiation protection rules, etc. 

Radiation Engineers in the field are mainly responsible for work planning, including determining 

internal and external monitoring requirements. The central technical support group provides 

procedure and technical basis development and oversight of the dosimetry, records, instrument 

calibration, and air sampling programs. For example, one manager indicated they used the 

central group to determine correction factors with respect to source strength difference for their 
 

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monitoring devices (e.g., dosimeters). In 1974, the RCTs joined the bargaining unit. INL now 

hires rent-a-techs to fill some of the radiological control staffing needs. 

Historically, workers at TRA and RWMC were often loaned to other facilities when needed. 

Sometime, workers themselves bid out to other facilities. With the transfer of the contract to 

CWI, this practice has stopped. 

According to the RadCon staff, refueling at NRF did not use any workers from other INL 

facilities. There were construction workers who provided support services at NRF. NRF was a 

very clean facility in terms of radiation protection. In fact, from an air sampling standpoint, it is 

cleaner inside of the facility than outside. 

Radiation Protection 

Historically, worker and management mentality indicated a lack of concern for safety issues. 

The field managers had very little control over this situation. The operations staff was often 

instructed by management to do whatever they could during their shift to get things done. In the 

past, the private construction contractors came, got their job done, and got out. They were 

always in a hurry to get things done and pushed the workforce hard. This attitude remained 

consistent through the various changes in contractors. 

Radiological control practices and policies were less stringent prior to the arrival of 

Westinghouse. After Westinghouse took over in 1984, they cleaned up most of the problems at 

ICPP. There was an improved contamination control program and better equipment was 

provided to field support. There was also a noticeable improvement in practices and worker 

knowledge of hazards in the workplace as training improved. 

As late as the early 1980s, there was a casual attitude toward RadCon. In the last several years, 

the pendulum has swung the other way, and controls are much more restrictive. For example, 

there is excessive use of Personnel Protective Equipment (PPE), and Radiation Work Permits 

(RWPs) are more specific to a particular job. There were several reasons for these improvements, 

including health and safety audits, public pressure, and lawsuits. The change in attitude may be 

partially due to the DOE Tiger Team audits. The Defense Nuclear Facility Safety Board 

(DNFSB) also came to Idaho for several visits in the 1980s and 1990s. Some site experts felt 

that DOE-Idaho did little to encourage improvement in the safety program, and that in fact, 

DOE-Idaho was contractor-owned. Some of the improvements in radiation safety are attributed 

to improved instrumentation and policies. Site experts also noted that there have been 

improvements in training, conduct of operations, and chemical sampling. 

From the 1950s to the 1970s, the ACL was 3,000 mrem/quarter. Dose was controlled on a 

weekly basis, with an allotted external dose of 300 mrem/week, or 600 mrem/2 weeks. These 

2 weeks could be sliding 2 weeks. With the signature of the Plant Manager, a worker could get 

up to 900 mrem for a particular job. In the 1980s, the ACL was lowered to 1,500 mrem per year. 

It was lowered further to 1,000 mrem/year, and most recently has been dropped to 

700 mrem/year. The goal was to maintain radiation exposure for an individual within 80% of the 

radiation limit. For some jobs, workers were authorized to receive 2-weeks exposure in 1 week. 
 

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The maximum allowable exposure for a given time was 900 mrem in 1 week. Limits could be 

reached within a matter of minutes for some jobs and required stay-time restrictions. Workers 

were typically reassigned to non-radiation work when they are bumping up against the ACL. 

Field Procedures 

In the 1950s, the INL site began to use Safety Work Permits (SWPs), which covered all aspects 

of safety concerns, including radiation, chemical, and industrial safety. Safety information 

provided on this form was of a more general type. Either the permit and/or the RCT 

communicated the Personnel Protective Equipment (PPE) requirements. 

In 1984 or 1985, the site switched to Radiation Work Permits: both job-specific and general 

RWPs. In the early days, the permits were primarily job-specific and had to be closed out on a 

daily basis. The general RWPs were for routine work with constant radiological conditions. 

Routine verification surveys were carried out to ensure radiological conditions remained 

unchanged. The job-specific RWPs were for work with changing radiological conditions. In this 

kind of RWP, pre-job survey, ongoing survey, and/or post-job surveys would be done. There 

were also ALARA calculations performed to estimate job-specific dose. RWPs were used for 

cell entries at ICPP and other high-risk jobs around the site. 

Initially the RCTs prepared the RWPs and determined the requirements. An RWP would be 

submitted to the foreman, the operations manager, and, lastly, the area manager for review and 

approval. Permits included PPE, job-specific dose limits, stay time, area dose rates, and other 

precautions as necessary. These requirements were determined based on process knowledge and 

experience. Dosimetry and bioassay requirements were not listed on RWPs until the 1990s. 

Currently, there is a more formal procedure for writing RWPs. Historically, the focus was on 

weekly limits. As a result, the original work permits did not have void limits. Radiological 

controls were based on past radiation and contamination surveys, field air sampling data, and 

former experience. RCTs would rely on pre-job air sampling to make adjustments to 

requirements for workers before entering an area. A RCT could initiate a 600 mrem limit (for 2 

weeks) for a worker with signatures from the shift foreman, shift manager, and senior manager. 

Radiation worker training was offered to employees at the site. The training addressed the 

different types of radiation; PPE; the concept of time, distance and shielding; and the proper way 

to don and doff personnel protective clothing. Annually, workers received refresher training on 

RWPs and PPE. There was a re-qualification process. There appears to be less radiation worker 

training now than before. 

Radiation Characterization 

At INL, the contaminants mainly consist of U, Pu, Cs, Am, Ce, and Sr. Uranium was associated 

with the dissolution and extraction processes at Building 601, SMC, ICPP, ATR, ARA, and 

SPERT areas. Plutonium was identified at many locations onsite, including ICPP, RMWC, 

ANL-W, the HFTF hot-cell facility, and the TRA. There were a number of places at the ICPP 

facility where plutonium and americium were identified, including A-cell, B-cell, C-cell, the 

Solvent Burner, and locations involved with the extraction process. There were trace amounts of 
 

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plutonium in recycled uranium received onsite. Mixed Activation Products and Mixed Fission 

Products were predominant radionuclides of concern onsite, especially Cs-137 and Sr-90/Y. 

These are found in facilities throughout the site. 

Radionuclides associated with particular processes included: 

• 	 At the test reactors, Co-60, Ir-192, and Cs-137 were the major concerns. 

• 	 The ROVER program processed graphite fuels with U-233/U-234. Uranium-233 was not 

uncommon in ORNL fuels process. 

• 	 Tritium was found at STRA, BDA, and in low levels at RWMC. 

• 	 At the ICPP radionuclides of concern include Sr-90/Y, Cs-137, Kr-85, Pu-238, Pu-239, Am241, iodine, and xenon. 

• 	 At Cell 603 at ICPP, the primary radionuclide of concern was Cs-137. 

• 	 Depleted uranium is of concern at SMC. 

• 	 Just recently, a vial of U-233 was found in an old building. 

• 	 There was/is soil contaminated with plutonium, strontium/tritium, and cesium. 

• 	 Waste transported during clean-up projects resulted in potential exposure to a number of 

radionuclides. 

• 	 Analysis of waste water from injection wells indicated the presence of tritium along with 

Sr-90. 

• 	 Tritium was found in the waste-water pond east of TRA. 

At the reactors, when sodium-clad fuel elements reacted with water, the fuel and fission products 

would leak to the water in the storage basin causing contamination spread. This would cause 

uptakes and chronic exposure problems for workers. General exposure rates were estimated at 

1 to 2 mR/hr in limited areas in the fuel storage and transfer areas. 

Although external exposure was the predominant radiation hazard, there were areas in the reactor 

that were contaminated as well; for example, there was contamination on the top of reactors and 

a lot of sealed capsules in aluminum tubes in the reflectors. There was a spider (basket) device 

in the end of each tube, so that coolant could go through. The capsules that held sample material 

would sometimes leak and cause contamination in the coolant water. The water was carefully 

monitored for unexpected contamination. Also, the tools used to manipulate items in the reactor 

pool were taken in and out of the reactor coolant water. In 1988/1989, there was a large release 

of tritium from the TRIGA Reactor in Building 604. 
 

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Both internal and external exposures were of concern at ICPP, with the former the primary 

concern. Pure alpha contamination was found at the Research and Development laboratories. 

The B-cell probably had the highest potential for internal exposure. Process Cell 601 was 

involved in the characterization and processing of uranium. ICPP had processed both enriched 

uranium and recycled uranium. Graphite fuels were also processed at ICPP. Fission products 

constituted the main radionuclides of concern and occurred in some areas with alpha emitters. 

There were in-process ambient sources like Cs-137. There were potential skin doses due to 

Sr-90/Y and Kr-85. At the calciner, the radionuclides of concern were Cs-137 for external 

exposure and Sr-90 and plutonium for internal exposures. 

Most of the fuel processed onsite was aged fuel of 1–3 years decay, with the exception of fuel 

associated with the RaLa process, which processed fuel immediately upon reactor discharge. 

The Pu-238 and fission products were extracted and separated from uranium, and the majority of 

the extracted uranium product was shipped to the Y-12 Plant. There were concentrated fission 

products and transuranics in the aqueous raffinate. Raffinates, a significant exposure hazard, 

were transported to 50,000-gallon underground storage tanks. Later, a calciner was built to 

granulate the wastes. At ICPP, there was an aqueous raffinate spill when a crack formed in the 

hydraulic fluid line, producing dose rates greater than 500 R/hour up to 15 ft away. 

Internal exposure was/is of primary concern at the RWMC. Pure alpha contamination was/is 

found at RWMC, due to disintegrated waste drums. There were/are plutonium products in the 

wastes from RFP. 

Tritium was a hazard in limited areas onsite. It was formed as a tertiary fission product in the 

high-enriched fuels. There were also tritium experiments onsite. At LTSF, drums were being 

retrieved from a carbon steel vault and caused a tritium issue. In the 1980s, there were tritium 

leaks in the gas plant. The operators had to wear bubble suits (not a routine procedure) to work 

in the area. Workers involved in this operation indicated that there were no recorded doses 

associated with tritium exposure. There is a still a trace amount of tritium in the surface soil in 

the gas plant. The site discharged tritium-related wastes 600 ft into the aquifer during the initial 

years of ICPP operation. Although tritium was present onsite, it did not constitute a significant 

source of exposure. 

Radiography 

Radiation generating devices include x-ray diffraction units, several accelerators, radiography 

sources, and calibration sources. Citrix was a neutron generator facility that had boundaries 

established to limit access to certain areas. Outside these boundaries, the dose rate was generally 

less than 5 mrem/hr. Inside the boundaries, the dose rate was in the mR/hr to R/hr range. 

Maintenance personnel would periodically modify experimental loops in the reactors and in 

cubicles where the cooling systems were located. These experimental loops were used to 

support Navy testing. Maintenance was also involved in repair of systems and reactor 

components. Radiographers would x-ray welds to insure they were intact. The process of 

welding and radiographing welds took a number of iterations. Portable x-ray machines and 

shielded Co-60 sources were used for radiography. Both the maintenance crew and the 
 

 

 

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radiographer often received significant doses from the radiation in the cubicles over a short 

period of time. Sometimes the workers would exceed their dose limit of 100 mrem/week. They 

then would need approval from management to allow them higher exposure to finish the job. 

During reactor shutdown, use of radiography sources was common. 

Radiological Contamination Control 

Survey and posting requirements were/are based on the existing radiological conditions (e.g., 

contamination level) and potential problems in an area. In the beginning, different color codings 

designated contamination areas; red, yellow, and blue. Different radiation areas were also 

established. Prior to the implementation of the complex-wide posting requirements, INL 

designated contamination areas as Zone 1, Zone 2, and Zone 3. Contamination limits for each 

zone were as follows. 

• 	 Zone 1: 200–5,000 dpm/100 cm2 β; 20–100 dpm/100 cm2 α 

• 	 Zone 2: 5,000–20,000 dpm/100 cm2 β; 100–2,000 dpm/100 cm2 α 

• 	 Zone 3: >20,000 dpm/100 cm2 β; >2,000/100 cm2

 dpm α 

Personnel Protective Equipment (PPE) and Respiratory Protection requirements were dependent 

on the radiological conditions of the work area. Single, double, and even triple anti-Cs 

(including coveralls, shoe covers, boots, etc.) were worn, depending on the level of 

contamination. Wearing multiple layers of Anti-Cs helped reduce beta exposure. Waterproof 

Anti-Cs were worn when working in wet areas, such as at reactor facilities. Bubble suits were 

used in the mid-1980s for the ROVER project. Some work locations were so contaminated that 

even three sets of PPE did not protect a worker from skin contamination. Zone 3 represented the 

most contaminated areas. 

• 	 Zone 1: Shoe covers, gloves, modesty clothing, single pair anti-Cs 

• 	 Zone 2: Shoe covers, two pairs of gloves, modesty clothing, double set of anti-Cs, no 

respirator 

• 	 Zone 3: Shoe covers, two pairs of gloves, modesty clothing, double set of anti-Cs, respiratory 

protection 

Later the criteria for determining PPE requirements changed. A single set of PPE was required 

in areas with 1,000 to 100,000 dpm/100 cm2 β/γ. A double set of PPE with respirators was 

required in areas with contamination levels larger than 100,000 dpm/100 cm2 β/γ. Plastics were 

used when liquid was present. Other safety factors, such as temperature, also influenced the PPE 

selected for a job. 

Respirators were often used as a preventative measure, and were issued to workers. The workers 

were taught how to maintain their respirators, and the respirators were inspected routinely. In 

later years, there was a single-use respirator policy for radiation and a multiple-use policy for 

chemicals. 

Respiratory protection ranged from half-mask respirators in the early days to airline respiratory 

protection for some of the hot cell jobs. Half-face respirators generally did not provide a tight 
 

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seal. The respiratory protection was upgraded from a half-face to full face if the protection 

factor with the half-face was not high enough. In the early 1980s, INL started to change from 

routine use of half-face respirators to full-face respirators. For highly contaminated areas such as 

hot cells, airline respirators or Scott Air Paks with compressed air supplies were used in place of 

a full-face respirator for particularly high-hazard work. 

At the reactors, workers and operators had to wear Anti-Cs, shoe covers, and gloves. 

Occasionally, respirators were used for some jobs. The bridge across from the entrance to the 

reactor top was normally roped off. 

Personnel Protective Equipment use did not always seem to be consistent between workers on 

the same job or over the course of time. For example, respiratory protection for ICPP hot jobs 

initially required a full-face respirator. Airline respiratory protection was eventually 

implemented with some jobs. The site eventually discontinued maintenance of the airline system 

and switched to Positive Air Purifying Respirators (PAPRs). No respiratory equipment was 

used for routine sampling. Process hot cell work initially required only a PAPR. This was 

upgraded to an airline respirator for grinding, cutting, or use of pneumatic tools in the mid-1980s. 

In the current work at RWMC, permanent staff may wear respirators; however, vendors do not. 

Historically, there were higher contamination levels in the work areas. Respirator equipment 

was not always used during sampling operations. In some situations, two workers in the same 

contamination area or high radiation area wore different levels and types of PPE. Management 

explained that there were different sets of rules that were applicable to each worker. 

Radiological Control staff indicated that RCTs observed every individual on a major job. 

Workers indicated that the RCTs at different facilities approached their work differently. Some 

RCTs were very thorough in their coverage, while others were laid-back. Starting in the 1980s, 

coverage for high-risk jobs, such as hot cell entries, was constant. Per procedure, RCTs were 

expected to ensure that individuals were wearing their dosimeter and to control exposures within 

approved limits. Dosimeter usage was sometimes difficult to police, since they were often worn 

under Anti-Cs. Remote operations and shielding were used to minimize dose on high-level jobs. 

Process sampling in the operation area was handled in fully shielded gloveboxes in a remote part 

of the facility. 

Self-survey was part of the doffing process for routine work. After finishing a job and prior to 

exiting, workers were required to change out their work clothes in the hot side and perform selfsurveys before they went over to the cold side. For egress in a high-contamination area (HCA), 

RCTs were required to perform frisking before the worker was allowed to pass through the 

Personnel Contamination Monitor (PCM). Pre-job surveys were done to access areas prior to the 

start of high-risk jobs. Objects leaving the area required survey by an RCT. When they became 

available at INL, individuals would pass through portal monitors before they exited a control 

area or left the plant site. 

Although eating, drinking, and smoking were not permitted in the immediate operations area or 

protected area, some site experts indicated that this was allowable in Radiation Buffer Areas. 

Operations maintained a coffee pot inside the control room, which was inside an RBA. The 
 

 

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control room was in the lower level of the building and was next to the sampling room. 

Operators carried water and food through the upper-level corridor beneath some leaky pipes. 

Then they had to walk down the steps to the lower level and go through a major operating 

corridor to get to the control room. 

Building 601/602 facility housed a cafeteria. Historically, workers were allowed to wear a lab 

coat over their working clothes and shoe covers during lunch to get to the cafeteria. This was no 

longer allowed as of the 1970s. The cafeteria was essentially surrounded by a contamination 

area. At one time, high contamination was spread all over Buildings 601 and 602, including to 

the cafeteria. A new cafeteria was eventually built in a separate building. 

Dressing rooms were the demarcation line between the cold and hot side of the facility. Workers 

were required to change their clothes and shoes and put on full Anti-Cs (particularly laboratory 

workers). Some site experts indicated that workers ate inside the change room. Some site 

experts used to eat on the “clean” side of the change rooms, immediately adjacent to the area 

where workers wore their plant clothing or frisked out of contamination areas. 

Historically, ICPP was very lax and contamination spread occurred on a daily basis. Some of the 

contamination spreads involved non-radiological areas. For example, in 1979/1980 management 

closed down the cafeteria due to a contamination spread all the way to the secretary station in the 

INL vice president’s office. They found the contamination level at the secretary’s chair reading 

10,000 dpm beta and gamma. As a result of this incident, portal monitors were installed. 

Site experts referred to the “ICPP shuffle,” which was the practice of shuffling in gravel to 

remove contamination from shoes. Machinists reported that chips of contaminated metals were 

often present in the CFA machine shop. 

Laundry services were available to pick up contaminated laundry and delivered clean laundry. 

Laundry workers used water-soluble bags, which could be placed directly into washers. A truck 

was used to transport laundry to and from the various onsite facilities. The truck was divided in 

half with the clean laundry on one site and the contaminated laundry on the other side. The 

contamination level on the laundry was generally low. The waste water was discharged to the 

leaching pond. 

In general, contamination control was based on providing the worker with PPE and 

decontaminating the affected areas whenever feasible. These methods of controlling 

contamination, however, were not always effective. Personnel and personal effects 

contamination was an issue at many facilities, particularly ICPP. Discussions with site experts 

indicate that there were numerous personnel contamination incidents. The incidents were 

usually associated with hot cell entries, evasive maintenance activities, or release of material 

from contained systems. In fact, a majority of the site experts involved in hands-on work had 

been involved in one or more contamination incident. In some cases, they lost their shoes, 

clothing, and even their hair. Although nasal smears were a part of the contamination control 

program, not all individuals with positive nasal smears were required to receive special wholebody counts (WBCs), urinalysis, and/or fecal samples. In fact, if the worker was able to remove 

the contamination, the incident was not even documented. 
 

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Personnel Decontamination 

When there was a personnel contamination incident, workers were required to scrub the 

contaminated area with soap and water; the RCT would have the individual scrub the affected 

area with abrasive material, which, in some cases, caused bleeding. Another method employed 

was putting tape over the contaminated area for the night. If the contamination was particularly 

fixed to the individual’s skin, as a final step, KMnO4 (the purple cow) was put on the area. 

Further scrubbing was sometimes necessary. There was some concern over the fact that the old 

decontamination room supervisor was mixing the solutions incorrectly. In the early days, there 

were no designated decontamination personnel at ICPP. Management would use different 

workers to perform decontamination. This approach was also used in other areas. Equipment 

was often decontaminated with the used of trichlorethylene (TCE). In the early years, if an 

individual got contaminated, they washed off the contamination as best they could and did not 

report it to keep themselves out of trouble. 

Air Sampling 

The purpose of air sampling at INL was to characterize the condition for workers, to verify 

posting, and determine the appropriate level of worker protection. Air sampling results were 

trended to identify low-level increases in concentration. An upward trend would result in an 

investigation of the cause. Most of the airborne radioactivity generated was the result of 

resuspension of contamination. As a result, contamination surveys were also trended to monitor 

for potential airborne hazards. 

The air monitoring program has included fixed air sampling at ICPP, job-specific air sampling, 

real-time air monitoring, and lapel air sampling. Continuous Air Monitors (CAMs) were used 

routinely from the 1950s forward for the detection of unexpected increases in airborne 

radioactivity in occupied area of most facilities. In 1981, job-specific air sampling was 

implemented (it was not used before). Breathing zone air sampling was not used often at the 

INL. There was no air sampling in high-contamination areas according to some site experts. 

The collection of krypton gas was performed for two to three batches every shift. There were 

CAMs placed in the facility to monitor the leaking. When there was enough in the air, the 

CAMs would sound the alarms. The operators would notify the RCTs and then continued 

working without stopping. The operators would wear bubble suits while bottling the krypton gas. 

Xenon gas was not collected, because there was no strategic value for it. There was no effective 

leak-testing of the bottles. The operators would just open the valve of the bottle to see whether it 

triggered the CAMs. 

Instrumentation 

One of the responsibilities of instrument technicians was to maintain radiation protection and 

criticality safety instrumentation. Instrumentation under their jurisdiction included PCMs, 

Radiation Area Monitors (RAMs), and Criticality Alarm Systems (CASs). Work involved 

calibration of RAMs and CASs. Collimated, directional point sources were used during the 
 

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calibration process. The CASs in Building 601, Building 640, and Building 602 are General 

Atomic systems. Those at the FAST Facility are made by Victoreen. When the technicians 

performed calibration, they normally used a collimated, directional point source. Neutron 

sources at FAST used cadmium as an absorber. At the ROVER facility, neutron sources were 

doped with poisons. 

Area Dosimetry 

Area dosimeters were placed in different locations to monitor work areas and perimeters. 

MicroR meter surveys were also performed in low potential areas to ensure postings were correct 

and to evaluate any trends. If there was an elevated dose rate, occupancy restrictions in the area 

were considered and the need for personnel monitoring re-evaluated. Although there may have 

been an absence of neutron monitoring in some areas of ICPP, there were wall-mounted neutron 

systems used in the corridors of some buildings. Also located in these areas were emergency 

dosimetry systems. 

External Monitoring 

Characterization of external exposure as acute or chronic is dependent on the work location and 

job assignments, and individual interpretation of the meaning of acute and chronic. In general, 

acute radiation exposures are more prevalent than chronic exposures. Radiation exposure is 

better characterized for some individuals as fractionated, where there were high exposures for 

short periods of time (well within established standards) followed by periods of no exposure at 

all. For example, ICPP had remote operations, but not remote maintenance, and, typically, there 

would be a lot of high-dose maintenance jobs over a short period of time. In some situations, 

workers could perform one to two jobs in a year and reach their allotted radiation exposure limit. 

There were some facilities at which exposure is better characterized as chronic (e.g., SMC, work 

outside hot cells). Most radiation doses at the reactors were due to exposures during shutdown 

activities. 

Maintenance and other personnel (health physics, operators, etc.) were involved in work leading 

to the receipt of higher total lifetime doses. For example, maintenance personnel interviewed 

indicated their cumulative exposures while working at the INL site ranged from 5–32 rem. Dose 

to maintenance workers has seen a large decrease over time. For example, a worker may have 

received 3–5 rem per year in the past, whereas his current annual dose may be approximately 

100 mrem per year. In some cases, personnel exceeded the limits of the time period. 

NRF, ANL-W, and INL used HSL dosimetry services initially. ANL-W has received 

radiobioassay, instrument calibration, and dosimetry services from HSL since the initiation of 

dosimetry services. In the early years, each entity used the same dosimeter. In the early 1960s, 

NRF became independent and eventually changed dosimeter types. 
 

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Calibrations 

For calibration, INL used a panoramic Cs-137 source starting in the early 1990s. The source was 

stored in a lead pig. The source also had an uncertainty of ±20% with <28% drift. Prior to this, 

DOE performed Cs-137 calibration using a beam irradiator and supplied the calibration to the 

contractors for a short period of time. Californium-252 was used as a neutron calibration source. 

The neutron albedo dosimeter was not calibrated with a neutron source. The TLD reader was 

calibrated using Cs-137 irradiated TLDs. The referenced Cf-252 neutron source used for the 

past few years replaces an older AmBe source, which was maintained by DOE in the mid-1990s. 

The neutron source TLD irradiations were used as a quality assurance check of the neutron TLD 

processing program. 

External Dosimetry Evaluation 

In 1998, a study was completed (EGG-1-98-04 or EDF/RDR-98005) to characterize neutron and 

beta-gamma fields in the various workplaces to verify that the monitoring badges were 

appropriate for the radiation fields. INL bought a multi-sphere from Chalk River and a BTI B-G 

filed spectrometer for use in this study. Dosimetry personnel spent 2 years studying the response 

of neutron and beta/gamma badges. In 1998, the study underwent an independent review by 

Idaho State University. The technical basis for the monitoring systems used at the INL have 

been documented in Technical Baseline for INEEL Personal γ-β Dosimetry (INEEL/EXT 01­

00636), Technical Basis of the INEL Personal Neutron Dosimeter (Report 960112), and 

Technical Baseline for INL 6776 Personal Neutron Dosimetry (RPT-131). Angular dependence 

was one of the items reviewed in these studies. Neutron spectral analysis was also included as a 

part of this study. 

There have been many studies to characterize the neutron spectra at INL, such as at the MTR, 

ETR, and ATR reactors. A Bonner Sphere spectrometer technique was used to document the 

field strength and to calculate the n-γ ratio. The 3” to 10” sphere ratio was used to determine the 

neutron field strength. 

Beta/Gamma Monitoring 

The issuance of dosimeters began immediately in 1951 and calibration records are available back 

to October 1951. The INL site benefited from the experience gained at other DOE sites. 

Dosimetry calibrations were performed by AEC Health and Safety Division, currently named 

Health Physics Instrumentation Laboratory (HPIL). The AEC was responsible for all the 

dosimetry equipment, calibration, bioassay, and environmental monitoring until January 1989, 

when contractors took responsibility for personnel dosimeter services. Effectively, the same 

individuals continued to perform the dosimetry service. One change was the requirement for 

calibration using National Institute of Standards and Technology (NIST) traceable sources. At 

times, this required that the specialized calibration services be provided by outside facilities, 

such as the Pacific Northwest National Laboratory. DOE served in more of a quality assurance 

role. The dosimetry program is currently in compliance with the requirements of the Department 

of Energy Laboratory Accreditation Program (DOELAP). The dosimetry systems had good 

sensitivity over the years. 
 

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Personnel may have worked at a number of different facilities within INL proper, such as ICPP, 

ETR, ATR, MTR, RWMC, and CFA. If an individual was assigned to more than one facility 

(e.g., maintenance, heavy equipment operators, etc.), the individual would have had a separate 

dosimeter for each facility. Regularly, badges were stored at the guard gates at each facility. 

Rack control badges were used to determine background dose rates. For the last couple of years, 

workers have been allowed to take their dosimeter home. However, most workers continue to 

leave their badges in the storage racks provided in the guard gates. Workers were told to wear 

their dosimeter on their chest, between the neck and waist, since the inception of the dosimetry 

program. In cases of potential partial-body exposure, RCTs could evaluate whether individuals 

needed to reposition their dosimeters. Historically, this was communicated as a part of radiation 

training. All badges were designed to wear or clip on the chest. The neutron dosimeter was 

combined with the standard dosimeter. 

Film badges and pocket ionization chambers (PICs) were used regularly by exempt support 

personnel associated with reactor operations. Prior to 10 years ago, a worker was required to 

wear a badge and a pocket dosimeter before going into a radiation area. Sometimes they also 

wore chirpers. For some period of time, the entire workforce wore dosimeters, including nonradiation workers, subcontractors, and construction workers. Currently, every worker entering a 

Radiation Buffer Area (RBA) is required to wear a TLD. 

Visitor, subcontract, and construction worker doses were maintained with the permanent 

dosimetry records. There was also a visitor form that allowed individuals to request their 

exposure record. Dose received by other DOE complex employees may have been sent to the 

particular DOE site of concern. At INL, contractors (companies) received monthly reports on 

individual worker doses received. Historically, the REASON or EXCUSE code in the dosimetry 

databases identified visitors. 

Neutron Monitoring 

From 1951 to the mid-1970s, neutron monitoring was performed mainly by using Kodak nuclear 

track emulsion Type A (NTA) film. NTA film was used for measuring fast neutron exposure, 

and has a detection limit of 10–14 mrem, with an upper threshold of 20 rem at a ±30% accuracy 

(documented). Thermal neutron exposure was not monitored until the thermoluminescent 

neutron dosimeter (TLND) was implemented, with the exception of some boron-lined PICs used 

at MTR. The results were not integrated into the “legal” (RESL) dose of record. In 1975, the 

albedo dosimeter replaced film. 

NTA was only effective for neutron energies of 0.5 MeV and above, and calculations were 

performed to estimate the total neutron dose. Sometimes, thermal neutron beams also had to be 

taken into consideration. Dale Hankins began his neutron dosimetry career at the MTR, 

published widely, and developed the multiple sphere Bonner Spectrometer system, including the 

3-in/9-in ratio technique for determining the average neutron energy. Hankins moved to LANL, 

but left the technology at INL. The neutron exposure attenuation was significant. To monitor 

the real-time neutron dose, a neutron-to-photon ratio was applied to the standard PIC reading. 
 

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These neutron-to-photon ratios were determined by field RadCon through survey instrumentation 

readings. 

Neutron monitoring devices were limited to specific areas of the site and not widely distributed 

to the workforce, as only a small percentage of the workers were involved in work resulting in 

neutron exposure. There were several areas where neutron surveys were performed routinely 

throughout the site. For example, neutron surveys were made on casks or drums shipped to or 

received from other sites. Many of the incoming casks or drums were filled with waste from the 

Rocky Flats Plant. On the average, the neutron dose rates were less than 10 mrem/hr. Neutron 

surveys were also conducted during reactor startup, around accelerators, during work with 

neutron-generating sources, and during work with fuel. Repackaging of plutonium nitrates 

normally produced very low neutron fields. Work with Cf-252 sources at TRA and ICPP 

required neutron dosimetry. 

Neutron monitoring was assigned to individuals working at the reactor areas, chemical 

laboratories, RWMC (starting 2004), the Calciner Facilities, the Process Equipment Waste Tank 

Farms area, and Citrix. Historically, no allowable threshold was implemented for neutron 

exposure. NTA film was routinely used at the reactor areas and at the chemical laboratories. In 

the early days, neutron exposure was primarily observed at the reactors. 

Some site experts believe the neutron monitoring program at INL has been inconsistent. For 

example, although the work has not changed, monitoring for neutron changed over the course of 

time. Workers noted that those outside the radiation boundary at RWMC do not participate in 

neutron monitoring, while those inside do. There are some disagreements between RadCon and 

other site experts as to whether neutron dosimetry was consistently used at ATR and ETR 

throughout the years. There was also some inconsistency between monitoring of permanent 

workers versus vendors, such as equipment handlers and excavators at RWMC. There was no 

routine neutron monitoring of some hands-on maintenance workers in ICPP. 

About 15% of the current workforce (about 700 people) is monitored by neutron dosimeters. 

The potential for neutron exposure primarily exists among personnel who work with neutron 

sources, with neutron beams at the reactors, at the dry storage spent fuel area, at RWMC, and 

with the pulsed neutron generator. There is some neutron exposure from work with plutonium. 

There was a potential for missed neutron dose in the early days, due to incomplete monitoring of 

the exposed population. For example, many laboratory analysts and chemists did not have 

neutron dosimeters, as they were not aware that there was an issue with neutrons. This lack of 

neutron monitoring could be verified by evaluating ambient neutron sources and cross comparing 

this information with dosimetry processing data. 

Real-time External Monitoring 

Pocket Ionization Chambers (PICs) and Electronic Dosimeters (EDs) were used in conjunction 

with film badges or TLDs for real-time monitoring. During high dose rate jobs, workers were 

given both a high-range and low-range PIC. The PICs had ranges of 200 mrem (TRA), 

600 mrem (low-range ICPP), and 1,500 mrem (high-range ICPP). During routine operations, 
 

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PICs were worn for up to 1 week. During high dose rate jobs, they would be used for the period 

of the job. Dose from PICs or EDs were used mainly for day-to-day tracking of worker dose to 

ensure regulatory compliance with dose limits. These units were also used in the estimation of 

dose when there were problems with film dosimeters (e.g., the film was completely black). If 

there was an unexpected exposure reflected by the PIC, the film badge would be pulled 

immediately for processing. 

Timekeeping was also used as a means of keeping track of and controlling real-time external 

exposure. Dose rates and stay times were used for high-risk jobs (e.g., entry into hot cells). 

RCTs were responsible for establishing area dose rates and ensuring individuals were in 

compliance with stay-time rules. The RCTs would also perform ongoing surveys in job areas. 

Now, electronic dosimetry systems are used to alert workers to unexpectedly high doses. EDs 

are set by radiological control personnel to alarm at a certain dose. Once the dosimeter reaches 

this dose, an alarm sounds and the worker is expected to exit the area. 

In the early days, pocket dosimeters were used and data collected by the Dosimetry group. 

These data were recorded together on the record sheets or cards. In 1958, film records were kept 

separately from pocket dosimetry data. The results from these are not available in the individual 

dosimetry files. RadCon field offices were responsible for assigning, reading, and maintaining 

PICs. The calibration check (they could not be adjusted, hence no calibration) of PICs was 

provided by Central Services. The pocket dosimeter or electronic dosimetry records from 1958 

to recent times were kept as field records and would be difficult to retrieve for comparison to 

film badge or TLD records. Since 1998, they have been kept in the same computerized database. 

Extremity Dosimetry and Multiple Badging 

Multiple badges and extremity dosimetry have been in use since 1953. Multiple badges included 

dosimetry for extremities (e.g., finger rings), the upper trunk, the lower trunk, and any other 

location deemed necessary by RadCon personnel. “Routine” badges were worn in addition to 

the multiple dosimeters provided for a job. Workers wore their primary dosimeters customarily 

on their chest or at their belt level. In the case of multiple dosimeters, the highest dosimeter 

value was recorded as the dose of record. The results from all badges of a multiple pack were 

maintained in the individual’s dosimetry file. The use of multiple badging was based on multiple 

high-level sources being present simultaneously in the work area, high-level point source with 

multiple workers, or other arrangements reflecting non-uniform fields. 

Formal procedures are now in place for extremity and multiple badging, including when and how 

to use multi-badging. The multi-badging records were kept in the regular worker files. The 

highest dose recorded was assigned as the dose of record up until 1995. At that time, INL 

implemented weighting factors to calculate dose from multiple dosimetry systems. 

There were situations, especially during maintenance activities, where there was partial-body 

exposure to workers. Many site experts were concerned about the potential in their jobs for 

extremity or nonuniform exposure. Multi-badging was used only in unusual conditions. For 

example, it was used during the SL-1 incident to monitor for high beta exposures. It was also 

sometimes used at the NWCF. Multi-badging was rarely used among maintenance and 
 

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operations workers. While typically there was some shielding afforded to the whole body during 

maintenance jobs, workers were often required to reach inside an area or around a pipe or valve 

where dose rates were much higher. In some cases, work was performed in tight spaces in close 

proximity to high-radiation sources. In the past, management required workers to perform some 

high radiation jobs without multi-badging. The multi-badging is very important in assessing 

external dose, as some jobs involved different dose rates at different body levels in a highradiation area. 

Prior to the 1980 time period, extremity and multi-badging were rare. Laboratory personnel 

often wore extremity monitoring. The jobs which typically required extremity dosimetry 

included the following: 

• Bottling of Krypton gas 

• Maintenance at the NWCF 

• Replacement of valve boxes in tank storage 

• Entry into hot cells (starting in the 1990s) 

Extremity dosimetry was implemented in the 1980s for crafts personnel. Not all operators were 

provided with extremity dosimetry in the earlier years. 

Hot particles were sometimes an issue in the reactor areas. These particles can be created by 

core internals change out, preventative maintenance on the reactor systems, and remote 

movement of reactor components. Although the fuel was nearly always under water when 

moved, they were transferred in air to the canal. This process generated hot particles, which 

deposited in the work area. 

Dosimetry Investigations 

If there is a lost or damaged dosimeter, a Personnel Exposure Questionnaire (PEQ) was 

completed. They were also completed for badges showing unexpected doses. The PEQs were 

sent to RESL to review and perform investigations in cooperation with the field. The 

investigations included a review of field radiological control data for the time period in question. 

In addition to field data for a particular individual, co-worker data was used to estimate dose 

when dosimeters were found to be overexposed. Real-time dosimeters of other workers on the 

job were also used for the purposes of maintaining real-time external exposure when the PIC or 

ED malfunctioned. “Doses of record” were changed accordingly. The resulting dose assignment 

was documented, including the method of dose estimation (e.g., based on field data or co-worker 

data). When the reason for spurious results could not be determined, a conservative dose was 

recorded. RESL would also perform trending studies on individual dosimetry records, on groups 

of individuals, and on whole facilities. 

Workers have expressed concern that the dose of record has underestimated the actual dose 

received. Workers consistently had access to their PIC results, and management periodically 

reported their film badge results to them. In later years, workers were issued an annual radiation 

exposure report. Numerous workers cited instances were their PIC went off-scale while working 

on a job. Workers did not specify which dosimeters went off scale. There were also a number of 
 

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instances reported of excessive film exposure (i.e., black film). These situations were common 

during ICPP cell entries, reactor shutdown, and in the old calciner facility. Several site experts 

reported that dose received on PICs was approximately three times higher than the dose reported 

in their dosimetry file [presumably deep dose] in some cases. One site expert recalled that he 

received 500 mR on his PIC, yet his film badge result was 100 mrem for the same period of time. 

Another individual had a dose of record of 3–4 rem; however, he believes the dose to be twice 

that much. Site experts indicated that the TLD seem to be more representative of the dose than 

film badges. 

Formal complaints have been made by workers regarding the lack of recorded dose on 

dosimeters. In these cases, RadCon investigated the situation by reviewing field records and 

interviewing workers. One possible explanation for discrepancies seen between PIC exposure 

results and dosimeter results is the substantial beta component in the field. It is not uncommon 

to see high energy β emitters causing PICs to read much higher levels. The same exposure 

detected by a dosimeter (i.e., film or TLD) would be broken down into penetrating and nonpenetrating dose. This beta dose would be reflected in the non-penetrating rather than 

penetrating dose. As a result, the whole-body dose would be smaller than the dose from the PIC. 

In areas where gamma exposure was the primary hazard, there was more agreement between the 

PIC and the deep dose from the film badge. At some sites, this phenomenon would also manifest 

itself in survey meters. A correction factor was developed for closed- and open-window 

readings as a result to correct for this situation. These correction factors are dependent on the 

assumed β energy. β-to-γ ratios can vary from 1-to-1 to as high as 25-to-1. Electronic 

dosimeters more closely matched the dosimeter results. 

There were no adjustments made to retrospective dosimetry records based on quality factor (QF) 

changes. Dose values were maintained consistent with the dosimetry standards at the time of 

measurement. 

Internal Monitoring 

Field or facility management personnel were always responsible for selecting individual bioassay 

programs. They were responsible for analyzing field results and identifying who to monitor, 

with the assistance of central radiological engineers with expertise in internal dosimetry, whether 

they were AEC/ERDA/DOE or the contractor. In the case of positive results, the internal 

dosimetry specialist would work with the field engineers to interpret the data and determine 

follow-up action. Internal Dosimetry was responsible for dose calculation and procedure 

development. 

The site implemented urine, fecal, and whole-body counting as a part of the internal monitoring 

program. The monitoring requirements have changed over time. In the 1950s to the early-1960s, 

there was a routine bioassay program for radiological workers. In early 1964, INL found that it 

was not cost effective to perform extensive bioassay monitoring on all workers. In the 1970s, 

there was no scheduled or routine bioassay program for any workers; rather, sampling was 

triggered by an incident. Several facilities included extensive urine, fecal, and whole-body 

counting bioassay programs that began in the late 1970s and continued until the mid-1990s. In 

the late 1990s, this was changed to a job-specific, event-based bioassay program coupling with 
 

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random sampling, which included fecal and urine sample analyses. Random sampling included 

workers performing high-risk jobs (e.g., 40 DAC/hour). For example, RCTs on high-exposure 

potential jobs would routinely be monitored. Records showed that any significant worker uptake 

would mainly relate to incidents. The event-based bioassay program considered risk factors like 

RWPs requirements, incidents, failed respiratory protection, exceeding the protection factor for 

the respiratory protection, and personnel contamination. Some baseline sampling was done for 

particular jobs. 

Internal monitoring was dependent on the work locations and even the period of operation. 

Radionuclides evaluated with the internal monitoring program included fission and activation 

products, uranium, tritium (limited), and plutonium in later years. This type of sampling was 

less common in the reactor areas. Routine programs typically included annual WBCs, including 

on employees from ANL-W. If there was a positive WBC, fecal and urine analysis were 

submitted. WBCs were also used for intake follow-up and after a significant contamination 

incident. During the period of routine bioassay, workers submitted up to several urine and fecal 

samples per year, depending on the area of the site. When there was a potential exposure to 

iodine, the site also performed thyroid counts. Some site experts believe that with the 

implementation of the RWP, radiological conditions improved and a more routine bioassay 

program was implemented. 

Site experts indicated they received the following internal monitoring, in practice, while at INL: 

• 	 In 1965, at TRA, there was routine urine analysis and WBC for workers. 

• 	 In the late 1970s or early 1980s, select personnel received semi-annual fecal and urine 

sampling with an annual WBC. 

• 	 In the mid-1980s, RWMC personnel received annual WBCs. 

• 	 Since the mid-1980s, ANL-W personnel have received annual WBCs. 

• 	 In the mid-1990s, personnel received an annual WBC and event-based in-vitro bioassay. 

• 	 Select personnel at the Special Manufacturing Capability (SMC) facility received 

depleted uranium urinalysis. 

• 	 Reactor personnel received only WBCs, and were not typically asked to submit urine or 

fecal samples. 

• 	 Support personnel, such as security guards and some crafts, indicated they only received 

a few WBCs while at INL, and were not placed on a routine monitoring program. 

Individuals who worked in process areas received bioassay and WBCs. Exempt support 

personnel (e.g., engineers) associated with reactor operations did not receive routine internal 

monitoring (WBC, fecal, or urine sampling). Construction workers received the same internal 

and external monitoring as did site workers. There was no routine bioassay program for nonradiological workers; however, the site would periodically include these individuals in bioassay 

monitoring to ascertain background information and validate the assumption of no internal 

exposure potential. 
 

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The internal exposure program relied on worker protection and contamination prevention, such 

as engineered design features, administrative controls, wearing respirators, using ALARA 

principles, and wearing detectors. 

In-Vitro Bioassay Sampling 

A urinalysis program for some radionuclides was implemented with the start of operations. In 

the mid-1960s, the urinalysis program was discontinued, since experience indicated that it was 

ineffective in discovering internal uptakes. The routine whole-body counting program was 

continued, while the bioassay program went to an incident-based protocol. In the mid-1980s, an 

internal exposure bioassay program was set up for detecting plutonium and Sr-90. There was a 

potential for missed dose prior to the implementation of these bioassay programs, as these 

radionuclides were present in the work areas. The current bioassay program is random and 

confirmatory in nature. It is a selection-based program that is dependent on the characteristics of 

a worker’s job. 

In the mid-1980s, the Radiobioassay Laboratory fell behind in analyzing samples for 

approximately 1 year. Follow-up samples for positive results would have been substantially 

delayed, resulting in potential missed dose if an intake occurred. At times, the workers reported 

a delay in incident-based bioassay. Sometimes, contaminated workers were sent to the wholebody counter after their contamination or radionuclide uptake was already gone from their bodies. 

There are about 1,600 to 2,000 in-vitro bioassay samples submitted per year. The numbers of 

positive results in more recent times have been low (10–30 positives per year). Because of 

recent remediation work at INL, the number of samples performed is expected to increase. 

Special In-Vitro Bioassay Sampling 

Field RadCon determined the need for special bioassay. “Triggers” (i.e., high air activity in the 

workplace, area contamination, and worker contamination) for special sampling were identified 

in operational procedures. The triggers varied over time. In 1979, if a worker in a contamination 

area had a nasal smear of 100 mrem/hr or more (a frequent occurrence), a bioassay would NOT 

be triggered. The potential missed (unrecorded) dose for a worker with this magnitude of uptake 

would be in the range of rems. No trigger levels were associated with field survey levels. 

Specific trigger levels for bioassay seemed to be in conflict with actual practice in the field. 

Over 50% of the site experts interviewed had been involved in a skin contamination incident 

(including some facial contamination incidents) at some time during their employment at INL. 

These incidents were especially prevalent in ICPP, but were not limited to this area. When asked 

if there were follow-up bioassays after these occurrences (which included positive nasal smears), 

many workers stated that no special bioassays were performed. 

In-Vivo Counting 

Historically, INL had an extensive in-vivo counting program. In 1958, the site implemented 

annual WBCs for radiation workers when the first whole-body counter came online. INL used a 
 

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mobile van, which traveled to work sites to perform counts. If the count was determined to be 

positive, the individual was sent to the fixed facility whole-body counter. As an example of the 

extent of the bioassay program, the site currently does 600–1,000 WBCs in a year. 

Thyroid counting was common in the late 1950s. This data must be evaluated with caution, 

however, because there were some uncertainties in the method used to calibrate the thyroid 

counters. 

Uranium 

SMC and ICPP were associated with depleted uranium and enriched uranium, respectively. 

ICPP ran a process to convert enriched uranium (uranyl nitrate), recovered from spent nuclear 

fuel, to UO3. Available information regarding operations at SMC was limited due to the 

classified nature of the work conducted there. The SMC area was primarily associated with a 

Type M depleted uranium form. For enriched uranium, there was an extensive bioassay program 

with fecal and urine sampling. The SMC facility did receive recycled uranium from other 

facilities. 

Two uranium background studies were conducted to determine the natural uranium 

concentration in urine. In the late 1980s, prior to the operation of SMC, a uranium background 

study was performed on a limited number of SMC workers. The background level was 

determined in the range of 0.04 to 0.33 μg/L. From this study, an action level of 0.25 μg/L was 

adopted. This study used a small population, however, and there was also some question about 

whether the workers monitored had been previously exposed to uranium. 

A second uranium background study was conducted using 16 individuals from the CFA. This 

represented a confidence level of 95% or 2σ (two standard deviations). These 16 people at CFA 

were used because they drank onsite water potentially contaminated with uranium. The 

background level was determined to be 0.045 ± 0.042 μg/L. From this study, an action level of 

0.175 μg/L was adopted. Offsite people were asked to participate in the study to obtain a more 

reliable background level, but none were willing to sign a liability waiver. This baseline 

sampling study was documented in EDF 32-56 (November 11, 2002). New possible confidence 

levels would be 99% (3σ) for 27 people, or 99.99% (4σ) for 1,000 people. 

Since plutonium operations were not a part of the INL site’s early mission, the plutonium 

bioassay program was not initiated with the startup of the site, but added at a later date. 

Operations with the potential for plutonium and other transuranic exposures included ICPP and 

RWMC. For example, plutonium was dissolved in the process stream, which periodically lost its 

integrity and barrier. Monitoring was originally designed to look for β and γ-emitters. In the 

1980s, the monitoring program began to evaluate bioassay samples for transuranics. The 

radiobioassay program has included analysis for both U-233 and Pu-238. Occasionally, there 

was a small uptake of Pu-238. The programs at RWMC (including the Advanced Mixed 

Transuranic Waste Treatment Program) presented a significant potential for uptake of 

transuranics. Transuranic waste stored in drums on the pads and in pits/trenches is currently 

being characterized, processed, and shipped offsite to the Waste Isolation Pilot Plant (WIPP) for 
 

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permanent disposal. Recycled uranium was processed at some facilities at INL. Plutonium, 

neptunium and technetium were in parts per billion concentrations. 

Some areas of the site contained high-fired uranium and plutonium oxide. The ROVER cell was 

used to process graphite space reactor fuel, resulting in the formation of high-fired uranium 

oxide. Some high-fired plutonium oxide may also have been shipped from Rocky Flats and 

disposed of at INL. At ICPP, carbonized fuel was processed such that high-fired oxides were 

formed. The fuel processed contained highly enriched uranium (~ 93% U-235). The recovery 

process resulted in a uranyl nitrate liquid product, which was converted to an oxide form for 

shipment. 

There were several studies on high-fired plutonium and uranium bioassay and lung counting. 

Studies of bioassay elimination curves for this material indicated a lung retention of 300 days. 

Fecal sampling was determined to be the bioassay method of choice for high-fired oxides. In the 

case of plutonium, a urine sample might only contain small concentrations for the first few days 

after a known intake. INL staff developed the procedures for detecting high-fired plutonium 

oxide in fecal samples. Some of the pioneer work for detecting isotopes of plutonium was 

conducted by the Health and Safety Division (and successor organizations). 

Tritium 

There was very little potential for exposure to tritium, with the exception of some periods during 

the 1970s. Tritium monitoring, consequently, was performed on a limited basis. The estimate of 

total population dose from tritium at INL is about 7,100 mrem. There are no known radon and 

thorium issues at INL above standard background levels; there was no thorium handled at INL. 

Internal Exposures 

Positive bioassay results were not uncommon in the past among support workers. Most of the 

exposure received by general service support personnel was from the ICPP facility. There was 

also some exposure received in the TRA and at the Power Burst Facility (PBF). The old calciner 

and old solvent burner had the reputation for having the highest internal and external dose 

potential at ICPP. In the 1980s, skin contamination and internal exposure were major problems 

for workers. 

There were a number of uptakes associated with operations at INL. Site experts collectively 

indicated that they had positive bioassay results for U-233, U-235, Pu-238, Co-60, and iodine. 

There was a general feeling that not all uptakes, particularly acute uptakes, were measured and 

documented in the individuals’ dosimetry files. 

Dose Assessment 

Most internal dose calculations have been documented. Doses were determined based on WBC 

data, urine bioassay, and fecal bioassay. Detectable internal doses almost always came as the 

result of an incident. The best fit method for the radionuclides identified in the bioassay was 

applied. Parameters resulting in the conservative dose estimates were used. In some cases, the 
 

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individual worker would provide several bioassays over a period of several months. The internal 

dosimetrist would then model the worker’s intake of radioactive material. The internal 

dosimetrist would also review RWPs written by the field engineers and other pertinent 

monitoring data when evaluating an internal exposure. 

No calculations of internal dose were completed with the use of air sampling data, except in 

unusual cases of dose reconstruction, where bioassay data was not available. There have been no 

comparison studies conducted between air sampling and bioassay-derived doses. The site did 

conduct a study (EDF 4510) on the use of Cs-137 as an indicator isotope for other radionuclides. 

A study completed at the Calciner with a cascade impactor determined particle sizes ranged from 

0.3 to 1 μm. The default particle size for internal dose calculations has been 1 micron AMAD. 

No recent particle size studies have been completed. 

As new data became available, internal dose from acute intakes were periodically recalculated. 

Both Annual Effective Dose Equivalent (AEDE) and Committed Effective Dose Equivalent 

(CEDE) can be found in the dosimetry records. Over the years, INL has recalculated doses and 

converted them to CEDE where data is available. The worker’s internal dose has been rolled up 

into their total lifetime doses, which include organ doses with weighting factors. The 

recalculation of internal doses often resulted in a reduction in dose, which was not adequately 

explained to the affected individuals. 

The INL Internal Dosimetry technical basis document includes missed dose calculations based 

on detection limits for various radionuclides of concern. An estimated intake and internal dose 

are calculated and provided in the document. 

Environmental Monitoring 

There has been environmental monitoring onsite since DOE (and its predecessors) began 

operations. Between 1951 and 1993, environmental monitoring was the responsibility of DOE 

and analysis was performed at the Health and Safety Division, followed by the Health Services 

Laboratory, and the Radiological and Environmental Services Laboratory (RESL). The onsite 

monitoring was taken over by the site management and operations contractor in 1993. There 

were both in-perimeter (onsite) and outside perimeter (offsite) monitoring stations for airborne 

and direct radiation measurement. The State of Idaho also has its own monitoring stations onsite 

and offsite. Some of these locations are co-located with the site-managed locations. The State 

sampling locations have been in place for approximately 15 years. Site personnel believe fenceline TLD measurements represent the best records for longer-worker impact evaluation. 

INL has collaborated with the National Oceanic and Atmospheric Administration (NOAA) from 

the beginning of the site operations. NOAA has provided the site with valuable meteorological 

data and analysis of atmosphere dispersion within the Snake River Plain. 

The primary operations associated with airborne releases at INL included the ANP project 

(specifically Initial Engine Testing); the ARA (which was the site of the SL-1 Accident); reactor 

operations at the Reactor Testing Complex (Materials Test Reactor, Engineering Test Reactor 
 

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and Advanced Test Reactor) and the Materials and Fuels Complex (Experimental Breeder 

Reactor II); the ICPP (including the Fluorinel and Fuel Storage facility and the RaLa program); 

the NRF; and several tests (such as the Fuel Element Burn Tests). The highest airborne releases 

occurred from the mid-1950s to the mid-1960s. Radionuclides released during operational and 

episodic releases included Cs-137, Sr-90/Y, I-129, I-131, I-132, I-133, Ar-41, Kr-85, tritium, 

transuranics, and other fission products. The RaLa processing runs at ICPP were the main 

source for I-131, but also resulted in the release of mixed fission products (such as Sr-90/Y and 

Cs-137). This was due to the very short decay time of the fuel prior to dissolution for 

reprocessing. The test reactor program resulted in releases of short-lived noble gases, such as 

Kr-88, Xe-138, Kr-87, Xe-135, and Ar-41. Measurable quantities of tritium have been released 

from TRA, NRF, and ICPP. Transuranic releases are associated with ICPP and RWMC. 

Cesium-137 and Sr-90/Y were released primarily as a result of fuel reprocessing. The 

radionuclide of most concern in relation to the NRF is Co-60; this isotope has been identified on 

the roof top. Roof-top contamination has not been limited to the NRF facility. 

ETR, MTR, and ATR had de-gassing tanks. At MTR, 100% of the coolants went through the 

degassing tank by using a vacuum. At ETR and ATR, smaller de-gassing tanks (about 

330 gallons) were used. They had to side-stream the coolants into the tank. Hydrogen and 

oxygen would be removed or degassed. Sometime, the gases would be blown out of the stack. 

Particulate releases at ICPP in the early 1990s were tracked by basic surveys of the facility 

surroundings. Contaminated areas from fallout were flagged and put into the release maps. 

Particulate releases were observed as a result of a new steam cleaning process of the ICPP stack. 

The material released was believed to be Cs-137 attached to white insulation material. 

Measurable radioactivity was associated with these releases. An air protection system now 

filters through a High Efficiency Particulate Air (HEPA) filter system, which has prevented 

reoccurrence of particulate emissions like those that occurred from the stack steam cleaning 

process. 

There are sources of surface water that flow onto the site, but no surface water flows off the site. 

Surface water percolates into the soil at playas located on the INL site, contributing to the 

recharge of the Snake River Plain aquifer. When reactors shut down, operators would purge the 

reactor system and fill the tank with clean water. The waste water was sent to a retention basin 

to allow for decay, and then discharged to an unlined pond, which allowed fluids to soak into the 

ground. 

The Snake River Plain was formed as a result of lava flows. There are fractures in the basalt 

leading down to the aquifer. Some transport of strontium, tritium, and iodine to the aquifer has 

been observed. There are also some detectable concentrations of tritium in the drinking water 

wells in the CFA. Although there is some groundwater contamination, the site has always met 

the drinking water standards for radioactive materials. The U.S. Geological Survey has 

conducted extensive studies of the aquifer and contaminant transport. 

The SL-1 incident caused the largest soil contamination of Cs-137 and Sr-90 at the ARA, with 

from 2–100 pCi/g Cs-137 after cleanup. The background level was 0.5 to 1.5 pCi/g Cs-137. 

Aerial surveys identify areas onsite with elevated soil concentrations. During decontamination 
 

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and decommissioning of the old calciner in 1994 or 1995, the soil in the area was found to be 

highly contaminated. As the workers dug into the ground, each layer of soil was hotter than the 

previous layer. Finally, they entombed the facility with concrete. 

Additional information on soil contamination (and the contemporary environmental monitoring 

program) is available in the Final Environmental Impact Statement for Waste Operations 

(ERDA-1536, September 1977). Since the initiation of Environmental Restoration, soil 

characterization data has become more accessible through the generation of the Environmental 

Data Warehouse Database. 

There have been several efforts completed to characterize historical releases at INL and resulting 

doses to offsite individuals. From 1989 to 1991, a study referred to as the Idaho National 

Engineering Laboratory Historical Dose Evaluation (HDE) was performed. S. Cohen and 

Associates (SC&A) and Risk Assessment Corporation (RAC) conducted an extensive records 

review, and generated a document summarizing the historical environmental releases and 

monitoring program at INL. This report contained information on historical airborne releases 

from INL facilities, provided some background on the monitoring program, and gave 

information on the environmental regulations applicable throughout the period. Information 

provided includes stack monitoring data, fence-line TLD data, release fraction data, and uptakes 

by vegetation, cow, and fish. 

The report identified that there were originally 26 meteorology monitoring towers (“met-towers”) 

at the INL site operated by the National Oceanic and Atmospheric Administration (NOAA). 

Now there are 33 met-towers in use. Before 1969, archival hourly data from one or more site 

meteorological stations were generally available back to 1954. Doses to offsite personnel, 

including a maximum hypothetical individual, from airborne emissions were calculated using a 

multiple puff model to estimate the potential ground concentration from release points. The 

assessments in this report are based on both onsite and offsite monitoring data. Where available, 

site data was verified with State environmental data collected from co-located sites. The report 

includes an analysis of various fuel element burn tests. It does not contain soil and water release 

data because these are insignificant pathways for offsite releases. Although valuable data is 

included in this report, it does not estimate potential onsite exposure to individuals. The final 

HDE report was issued in 1991. 

The Centers for Disease Control and Prevention (CDC) has been involved in a chemical and 

radiological dose reconstruction effort at INL since 1991. The Risk Assessment Corporation, on 

behalf of CDC, released Identification and Prioritization of Radionuclide Releases from the 

Idaho National Engineering and Environmental Laboratory in October 2002. 

Valuable resources for environmental data, containing soil and water release information, 

include the INEL Site Environmental Reports (DOE/ID-12082 series), quarterly and annual 

reports from the Health and Safety Division back to 1958, NESHAPS reports (DOE/NE-ID10890 (05) series), the 1995 Title V Operation Permit Application, INL Site Comprehensive 

Environmental Response, Compensation and Liability Act Closure documents, the 

Environmental Monitoring Program Report (INEEL/EXT-01-00447, September 2001), the 

Environmental Data Warehouse Database, and the Radioactive Waste Management Information 
 

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System (DOE/ID-10054, 1993). Annual and/or quarterly reports were generated by the Health 

and Safety Division starting in 1951. These reports include environmental monitoring data and 

are available back to 1958 through the environmental monitoring group. The database includes 

historical and new environmental data to support environmental restoration work. EG&G 

Nevada also conducted aerial surveys of the INL site. 

The State of Idaho has the Clean Air Act (CAA) primacy, except for National Emission 

Standards for Hazardous Air Pollutants (NESHAPs), Subpart H, National Emission Standards 

for Emissions of Radionuclides Other than Radon from Department of Energy Facilities. The 

Idaho Department of Environmental Quality conducts inspections of all regulated sources of air 

emissions. The Environmental Protection Agency requires air permits for radionuclide 

emissions, which stipulate High Efficiency Particulate Air (HEPA) filter testing, regular 

sampling of the warm waste pond, stack monitoring at the source, and continual ambient 

monitoring. The requirements in NESHAPs provide information on stack monitoring for 

specific radionuclides and their activities. The resulting site-wide dose from airborne emissions 

at INL is normally in the range of a few thousandths of a millirem (compared to the 10-mrem 

limit established in the regulations). Frenchman’s Cabin is used as the location where the 

hypothetical maximally exposed individual resides for the offsite dose calculations. The last 

operating permit application submitted by INL to the State was the 1995 Title V Operation 

Permit Application, which was updated in 2001. A number of Permit to Construct applications 

and modifications have been submitted for specific sources in the past 10 years. 

Medical Examinations 

The initial medical program at INL consisted of a physician, nurse, and aid-man, who were there 

to offer medical assistance to employees, if necessary. When operations at INL moved from 

construction to operation, the medical services were consolidated under the AEC. The original 

AEC guidance for Occupational Medical Programs was set forth in AEC Appendix 0528, 

Occupational Medical Program. Management surveys of the Medical Branch were conducted 

by the Medical Branch, Health and Safety Division as early as 1966. After the Energy Research 

and Development Administration (ERDA) took over from the AEC, the ERDA Manual 

Appendix 0528-A, Contractor Occupational Medical Program Handbook, was issued in 1975. 

In 1978, DOE Order 5480.8A, Contractor Occupational Medical Program, replaced the ERDA 

manual. The current requirements are found in DOE Order 440.1A, Worker Protection 

Management for DOE and Contractor Employees. In addition to this requirement, there were 

Occupational Safety and Health Administration standards (29 CFR 1910) requiring mandatory 

examinations (i.e., for metal workers, fissile material handlers, hoisters, and respirator users). 

The medical program was turned over to site contractors by the DOE in 1978. 

Medical examination requirements for workers have been based primarily on the physical work 

locations, materials an individual worked with (e.g., asbestos), and worker age. In the 1960s 

time frame, the frequency of exams was based on age and work location. Examinations were 

provided for individuals working with radiation, significant concentrations of toxic substances, 

or if there was a fitness-for-duty question. Medical Branch services were provided to all INL 

personnel. In the 1970s time frame, medical examinations were given to the whole work force. 

These exams continued for radiation workers up through the 1980s. Eventually, there were 
 

 

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changes made in the program due to the legal climate resulting in performing physical exams 

only when there was a reason for one. 

The following three excerpts from early AEC/ERDA Medical Services/Branch reports discuss 

exam schedules, and were described by the interviewees and provided to the SC&A team during 

the meeting: 

During 1960, ... a system to schedule laboratory work and physical examinations 

for all AEC and IDO contractor personnel was devised and placed into operation. 

All active medical records were reviewed to schedule employees for examination. 

Previously, the examinations had been scheduled either by one of the nurses in 

the plant areas or by request from the individual. This resulted in poor schedule 

control by the Medical Services Branch, and it was not possible to concentrate the 

examinations where the greatest benefit could be anticipated. .... The interval 

between complete physical examinations is dependent upon the employee’s work 

assignment, age, and health status. Unless the examining physician specifies a 

shorter interval, the following schedule has been employed: 

Age Radiation Area Employees Non-Radiation Area Employees 

(Badged) (Not Badged) 

18-24 4 years at age 30 

25-39 3 years 5 years 

40-49 2 years 3 years 

50-59 1 year 2 years 

60 and over 1 year 1 year 

(Source: 1960 Medical Services report by George L. Voelz, M.D., Branch Chief) 

The general schedule for physical examinations was revised during 1970 so that 

most employees would be examined at the same age. Prior to this, different 

schedules were applied for radiation and non-radiation areas. Experience has 

shown that with the level of radiation exposure at the NRTS, no increased 

incidence of medical problems is encountered in “radiation” workers. With the 

new schedule an employee may anticipate examination at time of hire, at age 25, 

30, 34, 37, 40, every two years until age 62 and then annually. ....”Lateral chest 

x-rays have been added to the routine PA every three years.” ... “A complete 

periodic physical examination at this time, in addition to the ‘routine’ physical 

and history, would include: chest x-ray, laboratory work – complete blood count, 

urinalysis and blood (chemistry) profile, pulmonary function testing, stool occult 

blood, PAP smear (optional), EKG and testing of visual and auditory acuity. 

Counseling and referral to private physicians follow in the event that 

abnormalities are identified. (Source: 1970 Medical Branch report by J.H. 

Spickard, M.D.). 
 

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Employee Health Examinations (1976) 

Physical examinations are performed to determine the health status and physical 

fitness of employees in an effort to aid in job placement and work safety and as a 

guide to strengthening the preventive medicine aspects of the occupational 

medical program where the health and welfare of the individual are concerned ... 

These physical examinations are of three types: 

a. Pre-employment: at time of hire 

b. Periodic: every year at age 45 and over and every two years if under 

45 years of age (at the conclusion of calendar year 1976, 63% of all fullservice INEL employees were under the age of 45 and 37% were age 45 or 

over. 

c. Termination: Upon termination of employment. 

The content of each type of physical examination is as follows: 

a. Pre-employment: Serology, Complete Blood Count, Urinalysis, Vital 

Capacity, X-Ray (PA & Lateral Chest), Audiogram, Orthorater (visual 

acuity, depth and color perception), height, weight, blood pressure, pulse, 

Electrocardiogram, physical examination. 

b. Long Periodic Physical*: In addition to items enumerated above for preemployment physical, the Long periodic physical contains: SMA 12/60 

blood chemistry profile, other lab tests deemed necessary by physician 

(e.g., GTT, Uric Acid, etc.) 

*Alternate Long and Short Periodic done on each scheduled physical 

c. Short Periodic Physical: Hematocrit, Hemoglobin and White Blood 

Count, Urinalysis, Audiogram, Orthorater, height, weight, blood pressure, 

pulse, physical examination and other lab tests deemed necessary by 

physician. 

d. Termination Physical: Hematocrit, Hemoglobin and White Blood Count, 

Urinalysis, Audiogram, Orthorater, Interview – nurse or physician. 

(Source: 1976 Annual Report – ERDA-ID Medical Division by John 

H. Spickard, M.D., Director) 

In the 1992/1993 time frame, the medical program ceased performing physical examinations on 

employees who did not have a mandatory driver for a physical exam, and started offering health 

profiles, including a laboratory blood chemistry profile, to those employees, in lieu of a physical 

exam. Employees whose work requirements involve medical certifications or health 

surveillances, as identified by their management, continued to receive physical examinations per 

those mandatory drivers, such as DOT Drivers, Firefighters, Fissile Material Handlers, Hoisting 

and Rigging Workers, Human Reliability Program (HRP) personnel, QA/NonDestructive 

Examination Inspectors, Reactor Operators, Respirator Users, Security Guards, Security Police 

Officers (SPOs), Asbestos workers, Beryllium Workers, Cadmium Workers, Hazardous Waste 

Operators, Hearing Conservation Program, Laser Operators, Lead Workers, etc. 
 

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Chest X-rays 

The technique used for chest x-rays prior to 1970 was a posterior-anterior view. Lateral and 

posterior-anterior view x-rays were given every 3 years from 1970 to approximately 1993. In 

2000, pre-placement x-rays were discontinued. INL did not use photofluorography, and no 

4 in x 5 in films have been identified by present medical staff. There have been no x-ray 

machines identified with this capacity. 

In the 1965/1966 time frame, there was a medical van installed with equipment owned by the 

INL site. In the late 1970s to 1990, surplus x-ray equipment, including that from the van, was 

either moved to other INL facilities or declared as surplus for use by other facilities. Current 

staff is not aware of the make and model of x-ray equipment used from 1949 through the late 

1970s. Current x-ray units were installed in 1990. 

In 1980, the Idaho Food and Drug Administration was responsible for inspecting equipment 

every 2 years. Their reviews included checks on the collimator, beam quality, radiation output, 

field alignment, source-to-image distance indication, and the entrance skin exposure. The 

administrative program was also reviewed. Preventative maintenance was responsible for 

calibrating, inspecting, and auditing equipment. This preventative maintenance was completed 

twice per year, and included mR/mAs and kVp/mAs tests. An offsite x-ray equipment 

maintenance contractor provides preventive maintenance support on the INL x-ray units. In 

1973, there was a test conducted by health physics personnel using a phantom, with TLDs placed 

on the surface and at a depth of 1.5 cm. Doses were determined for various techniques. The 

above documents are the primary sources available for occupational medical radiation exposure 

from x-rays. 

Incidents and Accidents 

In the earlier years, when an incident occurred, an Individual Event Report (IER) was generated. 

Later, the site recorded anomalies, personnel contamination, skin exposure, and over-exposure 

incidents on a Personnel Exposure Questionnaire (PEQ) form. For example, a lost dosimeter by 

a worker would trigger a PEQ process. PEQs were stored together and maintained with field 

RadCon records. Only the exposure assessment was maintained in the dosimetry record. This 

included primarily dosimetry information, such as organ doses, radioisotopes of concern, total 

doses for the events, and bioassay results. If there was a dose adjustment as a result of a PEQ 

evaluation, this was noted in the electronic dosimetry database. Anyone onsite could issue a 

PEQ, but they were usually generated from the Radiological Control Organization. A significant 

unplanned event would also be recorded in the Health Physics logbooks and on supporting 

surveys, air sampling, laboratory reports, etc. These records are stored with the field records. 

There were a number of recorded and unrecorded incidents or unexpected events that occurred at 

INL over the history of the facility. The recording practices for these incidents appeared 

inconsistent. Prior to the arrival of Westinghouse at INL, there was a high threshold for defining 

incidents; Westinghouse subsequently tightened controls. Site experts were not sure what the 

trigger levels were for incident reporting. A significant unplanned event would result in a formal 

incident report, critiques, and lessons-learned reports. Interviewees indicated that, on occasion, 
 

 

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personnel contamination incidents would seem to go unreported. Even when an incident or 

unusual event occurred, management did not effectively communicate the resulting internal 

and/or external doses to the workers. Eventually, individuals involved in an incident would 

receive a copy of the incident report 2 to 3 weeks after the events. In cases where no incident 

report was generated, this information was documented in the worker’s annual individual 

exposure report once a year. 

There were four major incidents at the INL site, including three criticality accidents and the SL-1 

reactor accident. The first criticality accident occurred in 1959 at the ICPP. While siphoning of 

uranyl nitrate solution in a waste receiving tank, there were multiple excursions due to favorable 

(for criticality) geometry. During the evacuation of the building, airborne fission products 

resulted in individual exposures. The second criticality accident, which took place in 1961, also 

occurred at ICPP; it resulted from improper configuration of uranyl nitrate solution in a vapor 

disengagement vessel. There was only minimal dose to personnel as the result of airborne 

fission products. The last criticality accident occurred in G-Cell at ICPP in 1978 as a result of 

excess material in the lower disengagement section of the scrubber column. There were notable 

releases out the stack. The SL-1 reactor accident occurred on January 3, 1961. As a result of a 

manual withdrawal of a control rod, the reactor went supercritical. A discussion of the SL-1 

incident is presented in a separate section. 

Following the 1978 G-Cell criticality accident, the plant was shut down for 8 days, and the 

dissolver had to be rebuilt. This dissolver was basically a garbage can that collected 

concentrated waste. The cleanup work required remote access sawing. Pipefitters had to reach 

deep into the contaminated lead line. When the piping was cut and dropped into a barrel, dose 

rates were so high that the RCT called for immediate exit from the area. The RCT surveyed 

areas around the dissolver bottom, and the instrument read off scale in excess of 50 R/hr. The 

RCT personnel got everyone out of the area and put temporary shielding materials around the 

garbage can. 

The ICPP facility was responsible for a lot of overexposure and personnel contamination 

incidents. This facility was originally designed as a pilot plant, but was put into routine 

operation. Work associated with the waste stream involved high beta dose fields. The β:γ ratio 

was in the range of 4:1, or even higher, in the calciner cell or the off-gas cell. Some of these 

incidents led to uptake of radionuclides. Several examples of non-routine exposures or unusual 

events at ICPP, mentioned by site experts, are provided below. 

• 	 A worker was working in a multicurie cell and was grossly contaminated. The HP 

personnel were able to decon the worker’s shoulders down to 200,000 dpm/100 cm2 . The 

area read 10 mR/hour. No bioassay sample was requested. 

• 	 One worker actually exited the plant with 365,000 dpm contamination on his knees, 

setting off the alarms at the gate. Eventually HP personnel went to the worker’s house to 

ensure no contamination was taken home. 

• 	 A mechanic was instructed not to touch a manipulator, but he inadvertently did. The 

RCT had to scrub him down to10 mR/hr. There was no bioassay done on this mechanic. 
 

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• 	 There was a gas bubble burst in the processing system and contamination was spread 

throughout the ICPP facility. There was an extensive cleanup effort. 

• 	 A pipefitter was leaning on a test train trying to remove it. The radiation meter went off 

scale. The test train was immediately replaced. As a result of this incident, the 

individual received his allotted dose. 

• 	 In the 603 Storage Building at ICPP, operators used buckets to transport contaminated 

waste water. 

• 	 In the tank farm, workers had to vacuum out and clean the facility due to a cyclone blow 

out (633). The workers had to work double shifts to get the cleanup done. 

• 	 In 1982, workers were dissolving Al-Zr fuel rods at ICPP. These fuel rods had to go 

through two separate solvents. They had problems with the mixture of O2 and H2. These 

fuel rods were about 40 ft long and were designed critically safe. This mixture problem 

caused an explosion so powerful as to lift the rods. 

• 	 There was a situation where a welder and machinist were burned out (i.e., exceeded dose 

limits) during work in the old calciner facility. This dose is not included in the lifetime 

dose, according to the affected individuals. 

• 	 Yellow uranium powder was found on the floor in E-cell. 

• 	 There were several incidents of positive nasal smears and facial contamination without 

follow-up bioassay. 

There was a significant release of I-131, I-132, and I-133 at the ICPP from the RaLa Project. 

Seven individuals inhaled the airborne iodine, with a maximum intake of approximately 40 μCi. 

These individuals were monitored with thyroid counts and urine sampling for several months. 

Detailed incident reports were generated, which were initially placed into individual dosimetry 

files. 

Routine maintenance of systems involved changing filters, which were later disposed of at 

RWMC. Dose rates around the filter and/or filter housing could be as high as 25 R/hour at 4 ft. 

As the Old Calciner Facility started to age, the cyclone started to leak and areas began to plug. 

This led to more personnel contamination incidents. Eventually, the Old Calciner Facility was 

shut down and the New Calciner Facility was put into operation. 

Leaks in system components or during transfers appeared to be a repetitive issue. One worker 

was sent to clean up waste water in the basement, because someone left the scum line open. 

There were significant α and β emitters associated with the task of reducing the water level 

accumulated in the basement. An intake of radioactive material resulted from this incident. 
 

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Trucks from TRA were loaded with waste materials and sent to the evaporator for processing. 

Operations personnel were responsible for unloading the material out of a tank via a pressure 

system. There was a lot of waste water spilled out on the ground in the unloading process, and, 

as a result, the ground became very hot and contaminated. Garden hoses were used to hose 

down the ground with service water. This spread the contamination around and the ground had 

to be cleaned. The waste water was absorbed into the surrounding soil. Sometimes, the tank 

would blow out gallons of waste materials. As this was supposed to be a contained system, no 

respiratory protection was worn during this operation. 

In the ICPP Calciner Facility, workers had to use friction saws to cut off valves that were 

contaminated with high levels of Cs-137, Sr-90/Y, and U-235. When they were drilling into the 

valves, sometimes filters were burned through, spilling contaminants, which would get all over 

their hands or faces. This often resulted in skin contamination. The airborne radioactivity level 

was also very high and may have been responsible for significant uptakes of radionuclides. 

In 1979, there was a worker uptake incident at the shift laboratory in the ICPP. This laboratory 

was responsible for analyzing samples from the uranium separation process and waste 

production at B627. The wastes produced were very high level. 

There was a time when the door flew open in the West Vent tunnel and caused a significant 

radiation release. In the tank farm, there were releases of materials to the soil with 20– 

30 mrem/hr. 

In the early 1980s, at the solvent burner, waste water dripped on to the ground from the pumps. 

There were releases of airborne contaminants out of the stack. These types of incidents had been 

occurring since the 1970s. During one operation, the pumps were primed too much and the 

process line had not caught up, causing too much solvent to build up inside the line. When the 

burner fired up, it exploded and excess solvents went up 100 ft in the process line and out of the 

stack. 

In late summer of 1981, one of the fans in the evaporator broke and plugged up the drain. Waste 

water had flooded the building. All operators were required to help decontaminate the area. For 

each of two entries per day, they would receive 300 to 400 mrem. The RCTs surveyed the area 

and found an average reading of 20 R/hr. When they got close to the drain, readings increased to 

50 R/hr. Standard PPE and a dosimeter were worn by personnel. The doses they received for 

this incident were recorded in their exposure records. Cesium-137 and Sr-90/Y intakes resulted 

for some workers. 

In about 1985, there was an excursion (incident) in the gas plant that blew up the bottom of a 

vessel storing krypton gas. All the CAMs went off in the plant. The operators ran into the 

facility to shut off all the valves. Later, by just looking at the instrumentation, they discovered 

that there was no krypton gas left in the system. Due to the absence of RadCon support, the 

operators were sent home. There was no bioassay sampling or contamination check. There were 

no doses reported or recorded for that incident. Critiques of the incident were held. The incident 

was not reported or recorded in the affected individuals’ annual medical and exposure files. 
 

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High radiation exposures were not uncommon during reactor shutdown, where an individual 

could burn out in less than 15 minutes. There were also hot particles found in the reactor areas. 

Although it was not the primary hazard, contamination was present as well. The reactor top had 

a particular issue with contamination. Several examples of non-routine exposures or unusual 

events at TRA, mentioned by site experts, are provided below. 

• 	 A worker was working on the reactor top at ETR when the Continuous Air Monitor 

(CAM) sounded. The resulting contamination required that the walls be scrubbed up to 

60 ft and the crane be decontaminated. 

• 	 A mechanic was working on a plug in the test train tube at MTR storage area. When he 

pulled the plug, his pocket dosimeter peaked and his chirper alarmed. 

• 	 A mechanic was reassigned to non-radiation work when his badge was observed to be 

black. 

• 	 The nozzle trench surrounding the core became contaminated when individuals worked 

in this trench. 

• 	 Workers at ATR were changing out the resin column and loaded the unit into a lead cask. 

They found the dose rates were close to 5 R/hour, forcing them to transfer the resin into 

another cask for disposal. They used a galvanized garbage can (55-gallon drum), put lead 

paper on the bottom, transferred the resin into the can, and then removed the can for 

disposal. 

• 	 A welder was assigned to re-weld a valve at the ATR. The stay time was limited to 15 

minutes. Upon exit, he noted his film badge was entirely blackened. The foreman sent 

him back in for an additional 15 minutes. 

• 	 A welder was making a cut in the in-pile tube to remove it and became contaminated. 

Area alarms were set off. 

• 	 In the 1960s, an experiment was transferred from ETR to a transport cask. The 

experiment dropped out of the bottom of the cask, setting off radiation area monitors. 

• 	 In 1964, a worker assigned to the TRA Hot Cell became contaminated and was later 

found to have detectable radioactive material in his left lung. 

• 	 In June 1984, there was a large spill at ATR. Workers were sent in to decontaminate the 

canal. It was so hot in the canal that a worker could not stay there for more than 

30 seconds. After the cleanup, the workers were found to have exceeded their dose limit 

for that month. They were reassigned to CFA initially, but then got assigned to work on 

another spill in another facility. 

• 	 There was a continuous release of argon from MTR while it was operating, because the 

graphite reflector was air cooled. 
 

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• 	 MTR/ETR had Ag-110m, Hf-181, Ta-182, Cr-51, and Hg-203 releases, which are typical 

of activation products at reactor experimental loop experiments. Rupture of experimental 

and sample lines resulted in spread of these products. 

• 	 There was a neutron chopper with a large neutron beam. A fire started in the paraffin at 

the MTR. 

In the 1964–1965 time frame, during a reactor shutdown at ATR, two welders went to weld the 

gas annular inside the reactor vessel. Some wooden boards were put on top of the pipes, so that 

the welders could slide in to do the welding. They were wearing 3 sets of anti-Cs, shoe covers, 

cotton coveralls, gloves, and half-face respirators, because it was extremely hot (radioactive) 

inside the vessel. Later on, while they were welding, they found they were all wet inside and out. 

It turned out that the reactor operators had been ramping up the reactor, without checking or 

realizing that the welders were still working in the reactor vessel. The water started to rise up in 

the vessel to be ready for the reactor fuel. The individuals were grossly contaminated and 

required extensive decontamination. This incident was not reflected in the involved personnel’s 

dosimetry files. 

Incidents also occurred at other facilities. In the late 1980s at the FAST facility, there was a 

contamination-spread issue involving antimony in dissolved fuels that spread all over the facility. 

Some fires associated with machining of depleted uranium occurred at the SMC facility. The 

latest event associated with an uptake occurred last summer at LOFTS. Some workers showed 

positive bioassay results for Sr-90/Y. The WBC results turned up negative. At the Custom 

Processing Facility, a glass vessel with plutonium and nitric acid exploded causing significant 

alpha contamination of a cell. One worker received a significant exposure as a result of the 

accident. There was a Type B investigation related to this incident. 

An individual received chelation therapy in the 1960s for an uptake of transuranics at the TRA 

facility. Union officials have documentation to support this. Another chelation, recorded in the 

medical files, occurred in 1993 after a person at ANL-W working with Am-241 in a glovebox 

was struck with the material after the flask broke. He was administered one very low chelation 

dose, based on TRU criteria, intended to calm him down. The INL dosimetry organization 

performs internal dose calculations to assist in the determination of when chelation is performed. 

Because ANL-W was under the auspices of DOE-Chicago, its processes were not known to the 

DOE-Idaho Occupational Medical Program at the time and, therefore, communication issues 

were a concern. These issues have now been rectified due to the recent contract consolidation. 

The PEQs are not provided to NIOSH as a part of the claimant packet. It is uncertain whether 

IERs were permanently preserved, or whether they are being provided to NIOSH. 

SL-1 Accident 

In January 1961, there was an excursion at the SL-1 reactor, which resulted in the death of three 

operators. The only evidence initial responders had that an excursion had taken place was very 

high dose rates. Initial responders included firemen, security guards, and radiation protection 
 

 

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personnel. Employees were exposed from the initial response, cleanup, and examination of 

accident personnel and reactor debris. 

At 8:00 AM on January 3, 1961, a fire alarm sounded in the SL-1 complex and the CFA fire 

station. A firefighting engine company of five firefighters and the Assistant Fire Chief 

responded to the SL-1 reactor building and looked around. The reactor was shutdown at the time. 

Eventually they made their way to the furnace room, where they found the activated alarm. 

There was no smoke or fire, so it was considered a false alarm. The fire alarm was reset and the 

fireman returned to the station. At 3:00 PM that same day, there was another fire alarm at the 

SL-1 complex. Firefighters went to investigate again and found the same fire alarm sounding 

next to the furnace. Again, there was no smoke or fire to be found and they reset the alarm. At 

this point, they requested that the alarm maintenance crew come to fix the alarm. They would 

not be available until the next day. The firemen reset the alarm and returned to the station once 

again. 

At 9:18 PM that night, a third fire alarm (generic alarm # 2-2-1 designated as the SL-1 complex) 

sounded at SL-1. At the time, it was -18°F outside and the firemen were not enthused about 

responding to yet another false alarm. They responded to the SL-1 site, where there was no 

guard present in the guard house at the time, and noticed there was an alarming personnel 

radiation monitor in the guard house. A phone call was made to the control room with no answer. 

It was unknown how many personnel were present on back shifts. The firemen then used their 

pass key to enter the complex. They first checked the alarm, which had given the previous false 

alarms. They found that this alarm was not sounding. 

The fireman entered the front door of the SL-1 reactor building to check the fire alarms in that 

area, as they thought the alarm may be coming from the reactor area. They went into the hallway 

and headed directly to the reactor control room. At that point, they noticed the dose rate in the 

control room area was 25 R/hour on the hand-held detection instrument they were using. 

Immediately, they saw the high-radiation lights were all flashing on the reactor control console. 

The three reactor operators were not in the control room and could not be found in the area. The 

firemen ignored all these signs. They were outfitted in Scott-Air Packs, standard fireman 

response gear, and had a radiation dose rate meter. They started up the stairway leading to the 

reactor floor. There were radiation alarms sounding and the meter they had went off scale, 

indicating a dose rate of greater then 250 R/hr when they were half way up the staircase. They 

thought the meter was broken, so they went to get another meter, which indicated the same 

results. They took one quick look inside the reactor floor, noting one person down a short 

distance from the entry door. At this point they knew something was terribly wrong. They 

retreated at this point and called the AEC dispatcher. 

A health physicist was requested, and the one available was at TRA, which was 18 miles away. 

It took him about 30 minutes to respond, as he had to find a vehicle and get high-reading 

instruments. When they went back to the reactor floor, they found debris and the first victim not 

far from the entrance. The dose rate at the entrance to the reactor was estimated at 6,000 R/hour. 

The first two victims were lying on the floor. The third victim was not immediately located, as 

he was pinned to the roof by a control rod. They knew they were in trouble, so they all got out of 

the building quickly. The Assistant Fire Chief immediately called the dispatching center. 
 

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A team of personnel that were in charge of operating the complex was sent out from Idaho Falls 

as a rescue operation team. The AEC recognized this as an emergency incident and authorized 

100 R for lifesaving and 25 R for saving equipment. The operators and co-workers volunteered 

to assist in the recovery of the bodies. The dose rates were so high that entry time was limited to 

a matter of seconds. The first rescue team was merely able to enter and lay a blanket on the floor, 

so that the second team could place the victims on a blanket and exit. The third team removed 

the victims to the outside, where radiation levels were much lower. The second individual was 

removed by plant workers from on top of the reactor several hours later. 

The site doctor responded from Idaho Falls in a government vehicle. The nurse responded to the 

scene in an ambulance from CFA. The first individual removed from the complex was placed in 

the ambulance with the nurse and was driven to Highway 20, where they were met by the site 

doctor. The victim was pronounced dead about 30 minutes after being retrieved. While 

transporting the first victim to meet the doctor, the nurse could not detect any vital signs, but she 

thought she saw movement in the victim. 

After the first two victims were out, they used a remote camera to look for the third victim, who 

was found pinned to the ceiling. A cherry picker equipped with a torch and a stretcher was later 

used to remove his body. The beams on the roof of the reactor were cut. The third victim was 

dropped into the stretcher and removed via a freight door to the outside of the building. This was 

done some time later. 

The victims were sent to ICPP and iced down in large sinks for preservation for further 

examination. This inadvertently led to some decontamination of the bodies. The Army doctors 

dismembered the bodies for analysis. This was the first time a human body had received such a 

high dose of radiation in peacetime. The bodies were eventually placed in caskets lined with 

lead, and their caskets were entombed in concrete for shielding purposes. No radiation could be 

detected outside the concrete. 

In addition to resulting in high external dose rates, there was a considerable amount of personnel 

and property contamination that resulted from the accident. The six firemen and others who 

initially responded to the accident were sent to a laboratory building 1.5 miles away. When they 

got there, the personnel monitor at the entrance was activated by their mere presence. They were 

so contaminated that they were asked to remove their clothes outside in the -18°F weather prior 

to entering the facility for decontamination. Their film badges were taken at the entrance for 

immediate processing. 

Following the SL-1 incident response, those involved were asked to submit daily samples (fecal 

and urine) for a 30-day period. After this time period, they submitted a urine sample twice a year 

and a fecal sample once a year. There were no explanations or debriefing by the management 

after the sampling or the incident. There were approximately 32 individuals involved in the 

initial response team. Among the initial response team were six firemen. 
 

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The ambulance was highly contaminated after transporting the victims. It was decontaminated 

and returned to the firemen for use at EBR-II. Eventually it was declared surplus and transferred 

to the local community of Blackfoot. 

The SL-1 reactor was allowed to sit for a few months after the recovery of the last body. At this 

point, a massive effort to remediate the site was undertaken by General Electric. The dose rates 

were still very high, though they had dropped since the initial recovery. All types of individuals, 

such as engineers, maintenance personnel, photographers, etc., were brought in to support the 

cleanup effort. There were over 1,000 individuals involved over the course of the cleanup 

operations. Operations were carefully planned, and remote handling was used wherever possible. 

As a result of the high beta doses, a belt with multiple dosimeters was used during the SL-1 

cleanup of the site. 

A team of GE engineers was sent to cut the roof off the building and remove the reactor core for 

analysis. The core was pulled and put into a transport cask. It was then taken to the TAN Hot 

Shop to be washed down, decontaminated, and examined. Criticality experiments were 

performed on the fuel to see if it would go critical if water was added. They bore a hole in the 

bottom of the vessel, causing particulate matter to be released. Control rods were placed through 

the top of the vessel. The reactor vessel bulged out as a result of the accident; it was hung in the 

Hot Shop for examination. The bulging of the reactor vessel was evaluated through remote 

measurements. Personnel removed the head of the vessel, took all the fuel out, and shipped the 

fuel to ICPP for processing. 

Medical personnel were aware that some workers had exceeded their dose limits prescribed for 

the time. One fireman responding that night received 18 R; however, this is not reflected in his 

radiation exposure reports. Other responders reported the same discrepancy in their dose records. 

No one they know of came to the Medical Office and complained about sickness due to these 

overdoses. Those firemen responding to the incident have not suffered a lot of adverse side 

effects. The nurse died of cancer a number of years later. As of 2005, of the six responding 

firemen, it is believed that three are still surviving today. 

High-Risk Jobs 

There were numerous higher-risk jobs associated with both routine and special operations at INL. 

These jobs were associated with high contamination, airborne radioactivity, and high radiation 

areas. It was not uncommon to encounter elevated radiation fields during maintenance and 

operations. Many of the operations onsite, especially those at the ICPP, required individuals to 

enter areas with very high dose rates. For example, at the New Calciner Facility, individuals 

worked in areas with 50 mR/hr to 50 R/hr dose rates. As a result of high exposure potential, stay 

time in these areas was limited. In some cases, the individual could receive the authorized limit 

for the job (up to 900 mrem with management approval) within a few minutes. Simpler jobs, 

such as tool placement, etc., would have to be performed by management, engineers, security, 

and other administrative-type personnel. The skilled craftsmen were saved to complete tasks that 

required special skill. Individuals were pulled from other areas of the plant, if necessary. 
 

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As a result of the high levels of contamination in the cells and the high dose rates associated with 

work in these cells, some of the highest doses received onsite were associated with the ICPP 

facility. The maintenance and repair of this facility required a variety of jobs, such as removal 

and replacement of the piping and filtration systems, welding, radiography, electrical work, 

changing bulbs, etc. 

Operations personnel were responsible for all kinds of tasks across the site, including work with 

waste streams. They would prepare areas for maintenance prior to the beginning of a job by 

performing decontamination, positioning lead shields, and/or cleaning up waste water as 

necessary. Operations were also responsible for collection of samples for analysis. At times, 

these samples were the result of process upsets. Other high-risk jobs mentioned by site experts 

included the following: 

• Filter change outs at facilities, particularly ICPP 

• Repair jobs at the calciner facilities 

• Process cell entries at ICPP 

• Shutdown activities at TRA 

• Maintenance work in the resin bed 

Many of the operations required were manual, rather than remote. Additionally, the original 

reactors were not well designed, which likely resulted in higher exposures than at ATR. 

Engineering controls, such as maintaining negative pressure or providing shielding, were utilized 

to minimize occupational exposure and release to the environment. The hot shops and change 

rooms maintained negative pressure. Radiation Work Permits were also used. Continuous Air 

Monitors (CAMs) and Radiation Area Monitors (RAMs) set up in different areas indicated any 

changes in radiological conditions. When either of these alarms sounded, individuals were told 

to exit the area immediately. 

The collection of krypton and xenon gas samples was a source of significant exposure to 

operations. The primary dose associated with this operation, assuming PPE was intact, was beta 

dose. Krypton escaped into the work place and through the stack as a result of leaks in valves 

and pipes. Workers had to use high pressure to push krypton gas into a cask or a bottle. About 

only 45%–50% krypton would be collected into the casks or bottles. One of the methods used to 

reduce β exposure at this and other facilities was wearing three pairs of anti-C clothing. There 

was a period of time when they placed plastic over the TLDs; the effect on the recorded shallow 

dose is not known. 

There were 6-in to 8-in pipes all over the basement room housing the calciner vessel. The 

radiation field was very high, in the range of 50 to 60 R/hr. Workers had to obtain special 

approval from management to work this job, because the expected dose was 5 rem. The vessel 

was very hot, in terms of both temperature and radiation. Temperature of 530°C made it difficult 

to work around the vessel. There was a spray head and a wing nut inside the bottom of vessel 

that required maintenance, but there was limited space to work in. In addition, there were oil and 

airborne contamination hazards around the vessel. The maintenance job involved reaching in for 

the wing nut, thus bringing the body close to the vessel; consequently, stay time was very limited. 

Those involved wore several layers of coveralls, anti-C clothing, gloves, etc. Personnel wore a 
 

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TLD, a pocket dosimeter, and extremity dosimetry. Individuals involved feel their record does 

not adequately reflect the exposures they received. 

In the mid-1980s, ICPP began the ROVER Project, which involved graphite fuels. As a result, 

graphite contaminants were found all over the facility. The primary concern was strictly α 

emitters. Worker protection required a special PPE and worker decontamination program. After 

worker decontamination and before exiting the facility, nasal swipes and then a WBC were done 

on the workers to ensure they were not contaminated. 

In about 1986/1987 time frame, an equipment operator was working in a subpile room at the 

ATR Facility. The task was to take out some pipe covering and lead shielding in the room 

underneath the reactor. The stay time was limited to 10 minutes, and the total job dose was 

900 mrem. Those involved wore a TLD and PIC on the chest. and two pairs of Anti-Cs. The 

exposure from this job was not reflected in the dosimetry record of personnel involved. 

In the early 1990s, maintenance personnel had to remove fuel and feed nozzles remotely in the 

New Calciner Facility. The removed nozzles were taken to be decontaminated and then to the 

hot machine shop to be cut apart. One of the kerosene nozzles in the calciner vessel was plugged 

up. They did not have the machinist to do that kind of job, so that they called the TRA machinist 

to do it. In this special task, no respirator was used. This was not a routine job and the individuals 

involved were not proficient in the operation. After the calciner stage, the calcine products were 

transported to the storage tank by piping. In this tank, there were dissolved stainless steel, 

cesium, and strontium. There were also slurries from the fuel dissolution, extraction, and denitration processes. The operators added nitrate, followed by organics, to the slurries in the 

process. They then added water to extract out uranium and plutonium. The denitration process 

was electrically heated and the products were reduced to powder. These powders were packed in 

sealed 2-kg cans or sealed overpacks. There was a lot of contamination around this facility, and 

personnel used paint to fixate the contamination. A 20-lb air purge was done at one point to 

remove air bubbles, during which a 25 R/hr field was measured around the system. Workers 

were wearing a single set of Anti-Cs and no respiratory protection. 

The Advanced Retrieval Program in the RMWC involves retrieving buried waste drums. Most 

of the waste drums in Pit 4, where they are currently working, are still in good shape. The 

airborne radioactivity levels are in the range of 15,000 to 50,000 DAC-hour. To contain this 

airborne material, a tent made of Herculite fabric was build over the pit. Drums in Pit 9 are in 

very bad shape and are expected to cause more of a radiological hazard. Rocky Flats wastes are 

stored in the buried waste drums being retrieved. 

Safety Analysis 

There were technical review standards and a “Safeguards Committee” that reviewed all 

operations in the early days. However, detailed written procedures were not produced. After 

ATR was built, extensive procedures for reactor operations were developed. Detailed written 

criticality safety procedures were not available until the later years, although criticality 

evaluations were performed. There were procedures in place to ensure criticalities did not occur. 
 

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For example, highly enriched uranium fuel elements were stored in wooden racks. Each rack 

was limited to six full assemblies. 

A Hazard Report was generated for MTR. This was the predecessor of the Safety Analysis 

Report (SAR). A Safeguards Report was generated for ETR in the mid-1950s. This is similar to 

a design basis document. An SAR was generated for ATR. 

Field Radiological Control Records 

A number of pertinent field radiological control records have been generated over the period of 

operation at INL. These records include Safe Work Permits (SWPs), Radiation Work Permits 

(RWPs), general area radiation and contamination surveys (with maps), job-specific radiation 

and contamination surveys, Health Physics/Radiological Control logbooks, air sample sheets, etc. 

The Health Physics/Radiological Control logs included information on the daily radiological 

activities and unusual occurrences. Field radiological control records were facility-based. For 

large generators, record types (e.g., RWPs) are stored together. For smaller generators, there 

may be multiple record types covering a particular period in the same box. 

All personnel contamination incidents were documented on a PEQ after their implementation, 

whether it was reportable or non-reportable. These were also used to document skin dose. 

RadCon maintained a PEQ file. Spread of contamination incidents were recorded in logbooks 

and on survey reports. 

Currently, radiological field records are stored in the RadCon office of the facility. Surveys, air 

samples, and other documents are reviewed routinely by the RCT foreman. These records are 

typically maintained at this area for less than 1 year. Like records are boxed and sent to the 

central radiological records area. Contents of boxes submitted are inventoried. The retention 

period for these records is 75 years. ANL-W records were maintained by ANL-W while they 

were under the jurisdiction of the University of Chicago. As many of the operations at SMC are 

classified, this facility maintains its own records. Field records are boxed separately from 

dosimetry records. 

Dosimetry Records 

AEC originally had the responsibility for maintaining dosimetry records. This provided some 

continuity between the numerous INL contractors. The reporting protocol for dosimetry data 

was based on the standards of the time. The initial standards used at INL were the AEC 0524 

Manual and the AEC 0525 Manual, which provided guidance on dosimetry and reporting. 

Subsequent revisions, including ERDA Chapter 0524, DOE Order 5480.1 Chapter XI, and DOE 

Order 5480.11, were the next standards that were implemented. The DOE Radiological Control 

Manual was implemented as mandatory guidance, followed by 10 CFR 835. The RadCon 

manual was required by contract up until about 5 years ago, and is currently a DOE Standard. 

There were code names used for dosimetry records, including INL site area codes, contractor 

codes, reason codes, and excuse codes. There is an official list of acronyms used across the site 

to refer to facilities, programs, etc. There was a historical dosimetry report for film badge usage 
 

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prepared by Foster Cipperly. This report documents information on the dosimetry and records 

program up to a certain time frame. It had been provided to NIOSH. 

Badge numbers were used as personal identifiers for some dosimetry records at the site. These 

badge numbers were not reused with one exception; the INL Site Managers/AEC/ERDA/DOE 

Field Office Managers/Presidents were always given the highest badge numbers (i.e., 1, 2 and 3). 

Dosimetry badge numbers were used for dose records and were unique. Dose records were 

stored under this number. 

Historically, INL categorized doses as penetrating (deep) dose, non-penetrating (shallow) dose, 

and neutron dose. In the early 1950s, beta-gamma (combined) dose was categorized as nonpenetrating/penetrating. Thereafter, the dosimetry algorithm called for separate categorization of 

non-penetrating and penetrating doses. Neutron dose was maintained and reported separately 

from the penetrating and non-penetrating dose. Current dosimetry records report “shallow” and 

“deep” doses, with the notation that penetrating deep doses would include neutron and gamma 

doses. Records provided in support of EEOICPA include “shallow” doses, “deep” doses, and 

“extremities” dose. Neutron doses were provided separately, but were also included in the 

“deep” doses. 

PIC and film badge results were initially recorded together on the same form. The dosimetry 

group eventually went to a logbook recording system, and the primary dosimeter results were 

maintained separately from PIC or electronic dosimeter results. 

Some bioassay records or results were maintained by the contractors during the early years, but 

incorporated into the records of the central dosimetry group. However, the AEC Health and 

Safety Division and its successors maintained the records of WBCs, bioassays, and internal dose 

evaluations that were conducted under their cognizance. In the late 1980s, the contractors took 

over the responsibility and continued with the existing records. 

Offsite exposure reports from personnel monitoring at other DOE complex facilities are received 

and integrated into the individual’s dosimetry file. There are several file cabinets full of 

exposure records received from other facilities. Offsite exposure data was supplied to NIOSH 

when available. 

NRF records were maintained by this facility and were not intermixed with other INL records. 

Some workers may have worked for both the DOE contractors and NRF facility. NRF doses are 

excluded from the compensation program, as they are associated with the Department of Defense. 

Individual dosimetry records were never destroyed, and there were no field records destroyed, 

with one notable exception. There was an incident at EBR-II where several boxes of records 

were contaminated. These records were thought to have been recopied and the originals stored 

or buried. There are subsets of records that have been removed from the site. When GE left the 

ANP, they took the worker exposure records with them. GE had a 4-digit number coding system 

for each individual worker. GE may have sent all the worker exposure records to ORNL. INL 

does have the exposure records for individual workers involved with SL-1, including those who 

were working for GE at that time. The information available in the INL files was mainly for 
 

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individuals involved with the SL-1 reactor examination (Hot Shop at TAN) and the cleanup. 

Personnel came from ORNL and other sites to support the SL-1 cleanup and went back to ORNL 

after they completed their tasks. SL-1-related exposure records for individuals involved in the 

cleanup work were kept in boxes stored in a room at the CFA 690 Building. 

As the DOE record policy has changed over the years, records may have been destroyed. For 

example, operational reports were disposed of after the established records retention period of 

the time. Personal dosimetry records had a retention period of 75 years and are intact. At INL, 

historical analytical data were thrown away at one point in time. They were reconstructed by 

memory and knowledge, in addition to personal files. Data was documented in summary and 

operational reports. 

Medical Records 

Existing staff have had the opportunity to review historical medical records. These records were 

not as extensive as the records maintained in more recent times. Without sufficient regulatory 

guidance in earlier years, each contractor employee given a medical exam had different 

requirements and medical forms. This included the particular elements of the exam. There were 

two flaws found in historical paper or microfilmed records. 

(1) Occupational medicine was not a specialty, and this type of information was not 

emphasized. 

(2)	 The quality of documentation was sometimes poor, due to the handwriting and 

subsequent microfilming of the original documents. 

These historical records have been kept in a system of records and are organized according to 

patient name. The electronic medical record system began in 1995. All digitized examination 

records include examinations and radiology interpretation (i.e., lateral, posterior-anterior, ankle, 

chest, etc.). 

Radiological control incidents were not included in the medical records unless medical 

intervention occurred. For example, if an individual was involved in an incident where a vial 

with I-131 was broken, there would be no mention of this in the medical record unless the 

individual received treatment. 

Unauthorized Practices 

Deliberate violation of RadCon requirements was considered a serious impropriety and 

management treated these allegations very seriously. If found guilty, an employee could be 

punished by discharge. 

In the 1950s and 1960s, some individuals would take off their dosimeters (to reduce the 

measured exposure readings), because they wanted overtime work. Conversely, there were other 

rumors that some workers would put their badges in high radiation areas to “cook” them, so that 

they would not have to work more hours in radiation areas. RCTs did not verify that individuals 

were wearing their dosimetry except prior to “hot jobs.” Research and evaluations into these 
 

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practices were conducted by facility management, but these rumors were not substantiated with 

facts. The RWP was the vehicle to control the work process. Historically, most people followed 

the rules, and badge misplacement was not commonly done deliberately. Sometimes the workers 

could have inadvertently moved the badge during work. INL procedure required a PEQ be 

completed when a worker went into a radiation area without proper badging. If individuals 

forgot their dosimeters, however, they would often not tell anyone, because they would get in 

trouble. 

In 1978, at the 603 Sludge Facility, workers were observed hiding badges in a lead pig. This 

incident was immediately reported to supervision and warnings were issued to those workers. In 

another incident, a person was found asleep in a contaminated area; he was fired as a result. 

In the beginning of the INL site, operational management allowed the RCTs to act as safety cops. 

Periodically, individuals were found eating, smoking, and chewing gum in radiologically 

controlled areas. Occasionally, RCTs would catch workers taking off their respirators in a 

contamination area to have a cigarette break. One time, a worker was found fishing in the waste 

water settling pond at TRA. His boat flipped over and the worker was up to his neck in waste 

water. If an individual was caught, an occurrence report was filed and disciplinary action was 

taken. 

Although ALARA was practiced at INL, supervisors knew they had an authorized “bank” of 

5 rem per year per worker, and considered this exposure as a resource for operational efficiency. 

EEOICPA Program 

The EEOICPA Program at INL began in July 2001. Initially, some data were not available or in 

place for claim support. INL has a Records Storage Facility, which currently contains 44,340 

boxes of records. Included are approximately 220 boxes of exposure/dose records that have now 

been entered into an electronic system, and approximately 1,400 boxes of personnel files that are 

accessed regularly for EEOICPA. Some of the ANL-W records were managed by the INL 

Records Program, including radiological dosimetry records. The INL Dosimetry Office decided 

not to use or scan the field records, because they were felt to be not required for the purposes of 

the EEOICPA Program. 

Dose records provided to NIOSH are submitted using batch numbers. The information provided 

includes dosimeter data for each badge cycle, in-vivo and in-vitro bioassay data, incident files, 

and a summary dosimetry report. Records provided to NIOSH are only redacted for Privacy Act 

reasons. For example, if there are records with more than one name on them, the record will be 

redacted, so that it contains only the information of the claimant. The EEOICPA Office includes 

medical and dosimetry records as requested by the Department of Labor (or as needed to help 

verify employment) for INL claimants. For the purpose of dose reconstruction, there is 

incomplete data and missing context of data during some periods of time. For those cases where 

medical records are incomplete or inaccurate, DOE has asked medical staff to make a guess of 

what the medical doses were. There is no technical basis for this guess. 
 

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Many workers shared by NRF (Naval Reactor Facility) and INL were construction-type workers. 

They were monitored by DOE and shared dose exposure data with NRF. NRF dose data were 

excluded from the data provided to NIOSH under the EEOICPA. The electronic database 

housing dose information has not been provided to NIOSH. Some claims contain over 2,000 

pages of exposure records. NIOSH has not requested and is not provided with field radiation 

records from INL. 

Documents associated with the SL-1 accident have been provided to the NIOSH Advisory Board 

on Radiation and Worker Health. Additional information on the SL-1 incident was published in 

1963 in the Health Physics Journal by the individual then in charge of radiation monitoring, 

environmental monitoring, and the dosimetry processing program. 

Release data from the Naval Reactor Facility were deemed classified until recently; therefore, the 

HDE report assumed values for environmental releases from this facility. In 2004, NRF 

provided environmental release and emission information to INL. NIOSH requested NRF 

environmental release documents from INL more than a year ago. These documents were 

redacted, but NIOSH has indicated no further interest in this data. The group within NIOSH 

requesting this data is unknown. INL has scanned and provided this data to the SC&A team. 

Radiological Control, DOE, and EEOICPA personnel interviewed believe these NIOSH 

individuals are very qualified to develop the INL Site Profile Technical Basis documents. 

INTREPID, a contractor to NIOSH, participated in interviews with the current INL and DOE 

radiological control staff and reviewed historical records. A number of documents were copied 

from the 1950s, 1960s, 1970s, and 1980s. The INTREPID team included technical experts with 

a long history of working at INL, which union officials believe to be a conflict of interest. 

ANL-W was initially included in the INL TBD; however, ANL-W now has its own TBD. 

Chemical Exposure 

In addition to being exposed to radiation, selected workers were also exposed to the following 

chemicals: 

• Beryllium 

• Asbestos 

• Cadmium 

• Lead 

• Mercury 

• Chromium 

• Iodine 

• Hydrofluoric acid from the denitrification process 

• Nitric acid 

• TCE 

At times, there was very little concern with chemical exposure. Workers were allowed to handle 

solvents barehanded. There was no safety control and no chemical monitoring, and workers 

often had no idea of what to do with chemicals. For example, historically at ICPP, if the acids 
 

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fumes were not overcoming individuals, the area was not monitored and PPE was not used. In 

the reactor areas, the primary concern was with radiation exposure. Engineers would approve 

engineering analysis of a job based on the acceptance limits of chemicals versus radiological 

hazards. The wet method for asbestos removal was not immediately implemented at the site. 

Brooms and dust pans were used to clean spills. Respirators were not worn. 

Asbestos blankets were used to control the spread of sparks from welding. They also were used 

as a kind of shielding. Instrument technicians, welders, electricians, machinists, maintenance 

workers, operators, and craftsmen often all worked in the same areas. They were using Inconal or 

Carpenter 20 (proprietary alloy) for piping to shield high temperature heat. Welders would build 

the shielding using these alloy materials. Among the other hazards at INL were the reactor 

beryllium reflectors, which had to be hand-filed in some cases. 

Miscellaneous 

• 	 Historical medical records are of little value in determining cancer effects from radiation. 

This is primarily due to the limited latency period for cancer since 1995. Medical records 

also do not document industrial hygiene issues effectively. In fact, many industrial 

hygienists took their records with them when they left the INL site. Similarly, many site 

employers took their records with them when they left. 

• 	 Medical personnel believe that facility releases were not a significant issue for onsite 

workers. Mortality rates have shown no excess for INL workers above background levels. 

• 	 Plant engineering support personnel were responsible for fixing things on the reactor, 

performing safety and core analysis, and developing concepts for new reactors. 

• 	 There have been formal audits performed through the operation of INL. Early audit 

reports were documented in operational reports. These reports were likely shipped to the 

Federal Records Center in Seattle. There were many audits performed on the dosimetry 

record system for the INL site. These audits were conducted by groups such as the DOE 

Tiger Team, DOE-ID, and DNFSB. 

• 	 There was incineration of low-level wastes at the Waste Experiment Reduction Facility in 

the later 1990s. Advanced Mixed Waste Treatment Program is not an incinerator. 

• 	 There are currently about 2,000 claims submitted from INL site workers. Many claims 

from ICPP personnel have been turned down so far by DOE and NIOSH. 

• 	 There was a bone cancer problem at TRA in the 1960s. 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

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ATTACHMENT 4: CONSISTENCY BETWEEN HANFORD, SRS, AND 

INL SITE PROFILES 

Table A-1: Occupational Medical Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory Savannah River Site Hanford 

Frequency of chest 

x-rays (Default) 

Before 1954 (PA): Annual 

1954–1969 (PA): New hires; 

Radworkers – 18–24 (every 

4th year); 25–39 (triennial); 

40–49 (biennial); over 50 

(annual); Non-radworkers: 

30–39 (every 5th year); 40–49 

(triennial); 50–59 (biennial); 

over 60 (annual). 

1970–1976 (PA+Lat): New 

hires; At age 25, 30, 34, 37, 

40; 40–62 (biennial); over 62 

(annual). 

1977–1978 (PA+Lat): New 

hires; less than 45 (biennial); 

over 45 (annual). 

1979 to 1/1990 (PA): New 

hires; high risk only (every 4th 

year). 

2/1990 to present (PA+Lat): 

New hires; high risk only 

(every 4th year). 

One annual x-ray procedure 

for each year or partial year. 

Posterior-Anterior View: 

Before 1946 – 1/1982: Preemployment, annual, and 

termination 

1/82–1/83: Pre-employment, 

annual, and termination for 

over 50 years; Biennially for 

40–49 years; Every third year 

for 39 years or younger. 

1/83–3/90: Biennially for over 

50 years; Every third year for 

40–49; and Every 5 years for 

39 years and younger. 

3/90 – present: Every 5 years 

Lateral chest x-rays also given 

periodically prior to 4/1997. 

Organ Dose 

Conversion Factors Obtained from ICRP 34 (1982) Obtained from ICRP 34 (1982) Obtained from ICRP 34 (1982) 

IREP Radiation Rate Not indicated in the TBD. Acute Acute 

IREP Radiation 

Type Not indicated in the TBD. Photons, 30–250 keV Photons, 30–250 keV 

IREP Dose 

Distribution Type Not indicated in the TBD. Constant Constant 

Total uncertainty 30% (no multiplier indicated) 30% (x-ray dose multiplied by 

1.3 and entered as a constant) 

30% (x-ray dose multiplied by 

1.3 and entered as a constant) 

No conversion factor 

indicated. 

Specific air kerma (mrad) for 

PA and Lateral used: 

Conversion Factor 

from PA to Lateral 

Pre-1954: 200 (PA) and 

200 (Lat); 2.5 2.5 

1954–1990: 52 (PA) and 

74 (Lat); 

1990–Present: 53 (PA) and 

76 (Lat). 
 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A-1: Occupational Medical Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory Savannah River Site Hanford 

Chest thickness not indicated. 

Distance accounts for body 

Chest Thickness thickness and space between 

person and film: 

PA View: 26 cm 

Lateral View: 34 cm 

PA View: N/A 

Lateral View: N/A 

PA View: 30 cm. 

Lateral View: 40 cm. 

Substitute dose 

conversion factors 

for thyroid, 

eye/brain, ovaries 

and analogues, 

testes, and uterus 

In the absence of collimation 

information, substitute DCFs 

used for the pre-1970 time 

frame (Table 4.0-1, Kathren 

2003). For skin doses, a 

backscatter factor is taken 

from NCRP 102 Table B-8. 

Substitute view and organ Use DCFs for lung for all 

DCFs applied to minimally other organs in thoracic cavity; 

collimated beams prior to for organs in abdomen, use 

1970. (Scalsky 2004, pg. 50) DCFs for the ovary (Scalsky 

2003, pg. 10) 

Analog organ for 

Thymus Lung Lung Lung 

Analog organ for 

Esophagus Lung Lung Lung 

Analog organ for 

Stomach Lung Lung Lung 

Analog organ for 

Bone Surface Lung Lung Lung 

Analog organ for 

Liver, gall bladder, Lung Lung Ovary 

spleen 

Analog organ for 

Remainder Organs 

None 

(Specific organ doses are 

provided for skin.) 

Lung Ovary 

Analog organ for 

Urinary/bladder and Ovary Ovary Ovary 

colon/rectum 

Analog organ for 

Eye/brain Thyroid Thyroid 

Posterior-Anterior View X-ray Techniques1,2 

Thyroid 

kVp: Unknown 

Site was not in operation. mAs: Unknown 

<1946 INEEL began operations in 

1949 (TBD#2, pg. 12). Site not in operation. SSD: 72” (183 cm) 

SID: 183 cm 

Filtration: 2.5 mm Al 

ESE: 120 mR 

2/1946 – 12/1950
 

Site was not in operation until 

1949. 

1949–1950: 

kVp: unknown kVp: 80 kVp: 80 

mAs: unknown mAs: 30 mAs: 25 

SSD: 153 cm SSD: 152 cm SSD: 72” (183 cm) 

SID: 183 cm SID: 183 cm SID: 183 cm 

Filtration: unknown Filtration: 1.5 mm Al Filtration: 2.5 mm Al 

ESE: 200 mR ESE: 108 mR ESE: 79 mR 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A-1: Occupational Medical Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory 

1951– 1954: 

Savannah River Site Hanford 

kVp: unknown 

mAs: unknown 

SSD: 153 cm 

1/1951–4/19/59 

SID: 183 cm 

Filtration: unknown 

ESE: 200 mR 

1954 on: 

kVp: 90 

mAs: 300 

SSD: 153 cm 

kVp: 80 

mAs: 30 

SSD: 152 cm 

SID: 183 cm 

Filtration: 

ESE: 108 mR 

kVp: 80 

mAs: 10 

SSD: 72” (183 cm) 

SID: 183 cm 

Filtration: 2.5 mm Al 

ESE: 79 mR 

SID: 183 cm 

Filtration: 2.5 mm Al total 

ESE: 52 mR 

4/1959 – 12/1970 

kVp: 90 

mAs: 300 

SSD: 153 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total 

ESE: 52 mR 

kVp: 80 

mAs: 30 

SSD: 152 cm 

SID: 183 cm 

Filtration: 3.5 mm Al 

ESE: 108 mR 

kVp: 80 

mAs: 10 

SSD: 72” (183 cm) 

SID: 183 cm 

Filtration: 2.5 mm Al 

ESE: 40 mR 

1/1971 – 1/1983 

kVp: 90 

mAs: 300 

SSD: 153 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total 

ESE: 52 mR 

kVp: 110–120 

mAs: 10 

SSD: 152 cm 

SID: 183 cm 

Filtration: 3.5 mm Al 

ESE: 44 mR 

kVp: 80 

mAs: 10 

SSD: 72” (183 cm) 

SID: 183 cm 

Filtration: 2.5 mm Al 

ESE: 40 mR 

1/1983 – 7/1985 

kVp: 90 

mAs: 300 

SSD: 153 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total 

ESE: 52 mR 

kVp: 110–120 

mAs: 10 

SSD: 152 cm 

SID: 183 cm 

Filtration: 3.5 mm Al 

ESE: 44 mR 

kVp: 100 

mAs: 10 

SSD: 72” (183 cm) 

SID: 183 cm 

Filtration: 2.5 mm Al 

ESE: 35 mR 

8/1985 – 3/1990 

kVp: 90 

mAs: 300 

SSD: 153 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total 

ESE: 52 mR 

kVp: 120 

mAs: 7.5 

SSD: 152 cm 

SID: 183 cm 

Filtration: 3.5 mm Al 

ESE: 33 mR 

kVp: 100 

mAs: 10 

SSD: 72” (183 cm) 

SID: 183 cm 

Filtration: 2.5 mm Al; 4.0 mm 

Al for CONX Type 12 

ESE: 35 mR 

3/1990 – 4/1997 

kVp: 100 

mAs: 300 

SSD: 153 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total 

ESE: 53 mR 

kVp: 120 

mAs: 7.5 

SSD: 152 cm 

SID: 183 cm 

Filtration: 3.5 mm Al 

ESE: 33 mR 

kVp: 110 

mAs: 6.7 

SSD: 72 “ (183 cm) 

SID: 183 cm 

Filtration: 4.0 mm Al 

ESE: 21 mR 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A-1: Occupational Medical Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory Savannah River Site Hanford 

kVp: 100 kVp: 120 kVp: 110 

mAs: 300 mAs: 7.5 mAs: 10 

4/1997 – 2/1998 SSD: 153 cm 

SID: 183 cm 

SSD: 152 cm 

SID: 183 cm 

SSD: 183 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total Filtration: 3.5 mm Al Filtration: 4.0 mm Al 

ESE: 53 mR ESE: 33 mR ESE: 17 mR 

kVp: 100 kVp: 120 kVp: 110 

mAs: 300 mAs: 7.5 mAs: 5 

2/1998 – 5/1999 SSD: 153 cm 

SID: 183 cm 

SSD: 152 cm 

SID: 183 cm 

SSD: 183 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total Filtration: 3.5 mm Al Filtration: 4.0 mm Al 

ESE: 53 mR ESE: 33 mR ESE: 11 mR 

kVp: 100 kVp: 120 kVp: 110 

mAs: 300 mAs: 7.5 mAs: 5 

5/1999 – present SSD: 153 cm 

SID: 183 cm 

SSD: 152 cm 

SID: 183 cm 

SSD: 183 cm 

SID: 183 cm 

Filtration: 2.5 mm Al total Filtration: 3.5 mm Al Filtration: 4.0 mm Al 

ESE: 53 mR ESE: 33 mR ESE: 11 mR 

Photofluorography 

kVp: 100 kVp: 80 to 100 kVp 

mAs: 60 mAs: not specified 

Technique Factors Photofluorography was not 

used at INEEL Site. 

SID: 102 cm 

Filtration: 2.5 mm Al 

SID: 102 cm 

Filtration: 2.5 mm Al 

ESE: ESE: 1.53 R 

Applies from 1951–1957 Applies 1945 to 1962 

1	 Refer to Scalsky 2004, pages 41–47 for SRS x-ray technique discussion. 2	 Refer to Scalsky 2003, page 18 for Hanford x-ray technique summary. 3	 N/A = not applicable; PA = posterior-anterior; LAT = lateral; kVp = kilovolt potential; mAs = milliampere-second; 

SSD = source-to-skin distance; SID = source-to-image distance; ESE = entrance skin exposure 
 

 

 

 

 

 

 

 

 

 

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Table A-2: External Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Missed Photon Dose 

Application 

Applies to workers with no 

recorded dose because they 

weren’t monitored or their 

results are unavailable; and 

workers who have a zero 

recorded dose (Rohrig, 2004, 

pg. 32). 

Applies to workers with no 

recorded dose because they 

weren’t monitored or their 

results are unavailable; and 

workers who have a zero 

recorded dose (Scalsky 2004, 

pg. 111). 

Applies to workers with no 

recorded dose because they 

weren’t monitored or their 

results are unavailable; and 

workers who have a zero 

recorded dose, (Fix 2004, 

pg. 75). 

Missed Photon Dose 

Methodology 

Divide the MRL by 2, and 

multiply by the number of 

zeros and not monitored 

periods (Rohrig 2004, pg. 32). 

Table 6B-1 (Rohrig 2004) 

provides potential maximum 

annual missed photon dose by 

period of use. 

(1) For a claimant-favorable 

maximum potential 

missed dose, use the limit 

of detection (LOD) 

multiplied by the number 

of zero doses 

(Scalsky 2004, pp. 111 

and 238) 

(2) Divide the LOD/2, and 

multiply by the number of 

zeros and not monitored 

periods; (Scalsky 2004, 

Divide the MDL by 2, and 

multiply by the number of 

zeros and not monitored 

periods (Fix 2004, pg. 75). 

Table 6E.6 (Fix 2004), 

provides potential maximum 

photon dose by year. 

pg. 242), or 

(3) Missed doses are added to 

measured doses and 

treated as a constant. 

IREP Dose 

Distribution Type 

for missed photon 

dose 

The percentage bias for gamma 

measurements at INL is plotted 

in Figure 6-9 (Rohrig 2004, 

pg. 22). The results lie within 

+27% to -43%. The TBD 

estimates that a realistic total 

uncertainty for photon 

dosimetry is about 35% at one 

sigma, which is consistent with 

the relative bias results. 

(pg. 34). 

(1) When using the LOD/2 

methodology, a lognormal 

distribution with a 

geometric standard 

deviation of 1.52 in 

Parameter 2 of the IREP 

input is used (Scalsky 

2004, pg. 116). 

(2) When simply adding the 

missed and measured 

dose, a constant is used. 

Lognormal distribution with a 

geometric standard deviation 

of 1.52.1

 The assessment at 

Hanford is based on the 

assumption that uncertainties 

from individual sources 

followed independent 

lognormal distributions. For 

each uncertainty source, a 

factor is assigned reflecting 

bias (B) and a 95% uncertainty 

factor (K); the uncertainty 

factor was determined so that 

the interval obtained by 

dividing and multiplying by 

this factor would include 95% 

of all observations (Fix 2004, 

pg. 27). 
 

 

 

 

 

 

 

 

 

 

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Table A-2: External Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Missed Neutron 

Dose Application 

Assign a missed neutron dose 

equivalent using Table 6B-2 

for the times when workers do 

not have reported neutron 

dose. For the period when 

NTA film was used, the dose 

should be multiplied by 1.25 

for all facilities except the 

MTR experimental floor or by 

2 for the MTR experimental 

floor when the MTR was 

operating between 1953 and 

1970. Then, the dose 

equivalent is apportioned into 

the IREP groups using Table 

6B-3. (Rohrig 2004, pg. 33). 

Assign a missed neutron dose 

if there is neutron monitoring 

between 1958 and 1962, if 

there is neutron monitoring in 

1971 or later, or there is 

indication of use of the 17 keV 

calibration curve for 

interpretation of beta/gamma 

film. Also applies to those 

who worked with Cf or Cm, 

maintenance workers, those 

involved in the PuAl target 

campaign, and those on routine 

plutonium bioassay. If the 

recorded neutron dose is 

greater than the calculated 

dose, the calculated dose is 

used (Neton 2003). 

Assign a missed neutron dose 

if the individual worked in a 

facility with a potential for 

neutron exposure. 

The vast majority of neutron 

dose to Hanford workers was 

received at the 200 West Area 

Plutonium Finishing Plant 

(PFP) facilities (pg. 74.) There 

is potential for significant 

missed dose in the 300 Area 

plutonium laboratory (308, 

309, 324), the 100 Area reactor 

facilities (i.e., reactors B, D, F, 

H, DR, C, KW, KE), the 300 

Area accelerator (3754B), the 

calibrations facilities (3745, 

318) and the Fast Flux Test 

Reactor (pg. 3). (Fix 2004). 

Missed Neutron 

Dose Methodology 

Use MRLs for neutron 

dosimeters because the neutron 

dosimeters were calibrated 

with neutron sources that had 

energies similar to those 

encountered in the workplace 

and most of the neutrons to 

which workers were normally 

exposed had energies greater 

than the 500 to 80-keV 

threshold of the NTA film 

dosimeters. 

Divide the MRL by 2, and 

multiply by the number of 

periods (week or month or 

quarter) (Rohrig 2004, pg. 32). 

Table 6B-2 (Rohrig 2004) 

provides potential maximum 

annual missed neutron dose by 

period of use. Multiplying 

factors are also applied here. 

A neutron-to-photon ratio is 

applied to missed and recorded 

photon dose for nonmonitored 

worker and workers with 

inadequate neutron monitoring 

(i.e., prior to 1971). The upper 

95% value is used for the 

maximizing technique. The 

geometric mean value is used 

for the best-fit technique 

(Scalsky 2004, pp. 240–241). 

After 1970, the assignment of 

missed dose is based on the 

limit of detection provided in 

Table E-10 (Scalsky 2004, 

pp. 241–242). 

It appears that an ICRP 60 

correction factor is applied to 

missed dose; however, this is 

unclear in the TBD (Scalsky 

2004, pg. 110). 

A neutron-to-photon ratio is 

applied to missed and recorded 

photon dose for nonmonitored 

worker and workers with 

inadequate neutron monitoring. 

The upper 95% value is used 

for the maximizing technique. 

The mean value is used for the 

best-fit technique (Fix 2004, 

pp. 75–77). 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A-2: External Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

IREP Dose 

Distribution Type 

for missed neutron 

dose 

No information on distribution 

with standard deviation is 

provided. 

Fractions of neutron dose 

equivalent were calculated. 

The conversion factors from 

dose equivalent to equivalent 

dose for INL spectra were also 

calculated. The ratios of 

average radiation weighting 

Lognormal distribution with a 

geometric standard deviation 

of 1.52.1 

Lognormal distribution with a 

geometric standard deviation 

of 1.52.1 

factor to average quality factor 

for the IREP energy groups 

showed variation. Default 

quality factor corrections were 

provided in Table 6-8 (Rohrig 

2004, pg. 32). 

IREP Exposure Rate No information is provided. 

Acute for beta and photon 

Chronic for neutron 

(Scalsky 2004, pp. 87 and 235, 

respectively). 

Acute for beta and photon 

Chronic for neutron 

(Fix 2004, pp. 8, 59, and 69, 

respectively) 

IREP Radiation 

Type (default) 

Photon, 30–250 keV 

Electron, > 15 keV, 

Neutron, 0.1–2, 2–10 MeV 

(Rohrig 2004, pp. 23 and 48) 

Photon, 30–250 keV 

Electron, > 15 keV, 

Neutron, 0.1–2 MeV 

(Scalsky 2004, pp. 49, 236, 

and 237, respectively) 

Photon, 30–250 keV 

Electron, > 15 keV 

Neutron, 0.1–2 MeV 

(Fix 2004, pg. 29) 

Organ dose 

conversion factor 

For photons prior to 1981, the 

conversion factor from 

exposure to organ dose should 

be used. For photons 1981 and 

after, the conversion factor 

from deep dose equivalent to 

organ dose should be used. For 

neutron doses in each IREP 

energy group, the conversion 

For the maximizing approach, 

a value of one is used (TBD, 

pg. 61). 

For the best-fit analysis, the 

dose conversion factors in the 

external dosimetry guide for 

the relevant exposure 

geometry. OCAS-IG-001 

Appendix A (NIOSH 2002) 

contains a detailed discussion 

of the conversion of measured 

The dose conversion factors 

for each, organ, radiation type, 

and energy ranged from 

OCAS-IG-001 are used. If the 

exposure geometry cannot be 

determined, default values are 

found in Table 6E-9 (Fix 2004, 

factors from ambient dose 

equivalent to organ dose for 

AP irradiation from Appendix 

B of NIOSH 2002 should be 

multiplied. (Rohrig 2004, 

pg. 34). 

dose to organ dose equivalent, 

and Appendix B contains the 

appropriate dose conversion 

factors (DCFs) for each organ, 

radiation type, and energy 

range based on the type of 

monitoring performed. 

(Scalsky 2004, pg. 242) 

pg. 77). No separate value is 

provided for the maximizing 

approach. 
 

 

 

 

 

 

 

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Table A-2: External Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Exposure geometry 

No information is provided for 

exposure geometry such as 

those for Hanford and SRS. 

For neutron exposure, it is 

assumed that the worker 

irradiation is in an AP 

geometry. (Rohrig 2004, 

pg. 34). 

Default exposure: 

Likely non-compensable 

workers - 100% AP 

Compensable worker – 50% 

AP, 50% ROT 

Compensable supervisor – 

50% AP, 50% ISO. 

Dose reconstructor has the 

option to choose the most 

appropriate exposure geometry 

for the individual. 

(Scalsky 2004, pg. 242) 

Default exposure: 

Likely non-compensable 

workers - 100% AP 

Compensable worker – 50% 

AP, 50% ROT 

Compensable supervisor – 

50% AP, 50% ISO. 

(Fix 2004, pg. 77) 

Photon Adjustment 

Factors 

(Recorded Dose) 

No adjustment for photons is 

provided. 

Multiply by 1.119 for years 

prior to 1987. Multiply by 

1.039 for 1987. No adjustment 

is needed post-1987 (Scalsky 

2004, pg. 238). 

Note: Taylor et al.(1995) 

indicates that the 1.119 

adjustment factor should be 

No adjustment for the multielement dosimeter, TLD, or 

gamma dose. 

For 200 Area plutonium 

workers prior to 1957, the 20% 

of the open window dose is 

applied through 1985 and the 

1.039 adjustment factor should 

be applied for 1986. No 

correction is required for 1987 

and after. 

added to the penetrating dose 

(Fix 2004, pg. 73). 

IREP Dose 

Distribution Type 

for recorded photon 

dose 

No information is provided for 

IREP dose distribution type. 

Constant. The adjustment 

factor encompasses the 

uncertainty so no additional 

uncertainty factors are 

included. 1 

Constant.1 

Recorded Neutron 

Dose Adjustment 

Factor 

(Prior to 1971 – 

SRS; Prior to 1972 

Hanford) 

Table 6-8 lists the 

recommended default values 

for the dose equivalent 

fractions and quality factor 

corrections for adjusting 

recorded dose (Rohrig 2004, 

pg. 31). 

NTA film is considered 

inadequate for use in dose 

reconstruction due to the 

energy dependence. The 

missed neutron dose approach 

is applied for this period of 

time. If the measured dose 

from the NTA is greater than 

the calculated dose, this value 

is used and the ICRP 60 

conversion factor is applied 

(Scalsky 2004, pg. 238). 

NTA film is considered 

inadequate for use in dose 

reconstruction due to the 

energy dependence. The 

missed neutron dose approach 

is applied for this period of 

time (Fix 2004, pg. 48). 
 

 

 

 

 

 

 

 

 

 

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Table A-2: External Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Recorded Neutron 

Dose Adjustment 

Factor (7/78–12/83) 

See above. 

In order to calculate the dose 

input for the IREP, Table E-1, 

the recorded neutron dose must 

be separated into neutron 

energy groups as shown in 

Table E-3 and subsequently 

converted to ICRP 60 (1990) 

methodology (Scalsky 2004, 

235–238). 

When using the four-chip 

HMPD during the period of its 

use from July 1978 through 

December 31, 1983 in Hanford 

200 and 300 Area plutonium 

facilities only, multiply the 

recorded neutron dose by 1.35. 

At all other times, divide the 

dose into the facility specific 

neutron energy bins, and 

multiply by the ICRP 60 

conversion factor (Fix 2004, 

pg. 74). 

Recorded Neutron 

Dose Adjustment 

Factor (1/72–6/78, 

1/84 – present) 

See above. 

In order to calculate the dose 

input for the IREP, Table E-1, 

the recorded neutron dose must 

be separated into neutron 

energy groups as shown in 

Table E-3 and subsequently 

converted to ICRP 60 (1990) 

methodology (Scalsky 2004, 

235–238). 

Divide the recorded neutron 

dose into the facility specific 

neutron energy bins, and 

multiply by the ICRP 60 

conversion factor (Fix 2004, 

pg 74). 

IREP Dose 

Distribution Type 

for recorded neutron 

dose 

Various for different IREP 

energy groups. See Table 6-8 

(Rohrig 2004, pg. 32). 

Constant. The adjustment 

factor encompasses the 

uncertainty so no additional 

uncertainty factors are 

included. 1 

Constant1 

Shallow Dose 

Adjustment Factors 

Shallow dose adjustment 

factors are not addressed in 

this TBD. 

Shallow dose adjustments 

factors are not addressed in the 

TBD or SRS TIBs. 

Shallow dose adjustments 

factors are not addressed in the 

TBD. 

Low-energy photons 

(< 30 keV) 

INL reported doses as 

penetrating and nonpenetrating. The penetrating 

dose corresponds to the deep 

dose equivalent and the 

nonpenetrating dose plus the 

penetrating dose corresponds 

to the shallow dose equivalent. 

1954–1981 Subtract the 

reported deep dose from the 

shallow dose for plutonium 

workers. 

1982–present. Plutonium 

workers are those individuals 

that worked in 321M, 221H – 

B line, 221F – B line, 772F, 

235F, 773A, 736A, and other 

plutonium storage areas (Neton 

2004). 

(For testicular, breast, or skin 

cancer) 

The stated Hanford practice to 

include 1/5 of the shallow dose 

based on a 16-keV calibration 

to the deep dose for Hanford 

plutonium facilities workers 

could resolve this source of 

potential under-response 

around 17 keV (Fix 2004, 

pg. 26). 

For 200 Area workers prior to 

1957, the 20% of the open 

window dose is added to the 

penetrating dose, (pg. 14). 
 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A-2: External Exposure Default Assumption Comparison for INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

IREP Dose 

Distribution Type 

for recorded shallow 

dose 

Not included in the TBD. 

Shallow dose is addressed 

from a technical perspective in 

the TBD, but no direction is 

provided to the dose 

reconstructor (Scalsky 2004, 

pg. 97). 

Not included in the TBD. 

IREP Radiation 

Type for recorded 

dose 

Specific to the particular 

facility for beta, photon, and 

neutron dose. For example, in 

the reactor area 100% of the 

beta doses is assumed to be 

>15 keV, 75% of the photon 

dose is >250 keV, and 25% of 

the photon dose is 30–250 keV 

(Rohrig 2004, pg. 23). 

Specific to the particular 

facility for beta, photon, and 

neutron dose. For example, in 

the reactor area 100% of the 

beta doses is assumed to be 

>15 keV, 50% of the photon 

dose is >250 keV, and 50% of 

the photon dose is 30–250 keV 

(Scalsky 2004, pg. 98). 

Specific to the particular 

facility for beta, photon, and 

neutron dose. For example, in 

the reactor area 100% of the 

beta doses is assumed to be 

>15 keV, 75% of the photon 

dose is >250 keV, and 25% of 

the photon dose is 30–250 keV 

(Fix 2004, pg. 29). 

1

 These parameters were obtained from review of several dose reconstruction IREP input sheets. 
 

 

 

 

 

 

 

 

 

 

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Table A-3: Comparison of Default Assumptions for Internal Exposure at INL, 

Savannah River Site, and Hanford
 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Particles Size 

(default) 

5 micron AMAD (default) for 

RWMC (TBD, pg. 34). 2.4 

micron AMAD (default) for 

SMC (TBD, pg. 29). No 

default values are provided for 

other INEEL facilities. 

5 micron (Scalsky 2004, 

Section 4.0, Attachment D) 

5 micron 

(Bihl 2004, pg. D-10) 

Intake Type (default) Chronic (TBD, Table 5.7-1). Chronic (Scalsky 2004, Section 

4.0, Attachment D) 

Chronic 

(Bihl 2004, pp. 7–9) 

Default Excretion 

Volume Not included in the TBD. 

1.4 liters/day 

(Volumes less than 1.4 

liters/day are corrected by 

normalizing the actual volume 

to 1.4 liters/day. Samples 

recorded as activity per 1.5 

liters are not corrected.) 

(Scalsky 2004, pg. 70) 

Uses a urinary excretions value 

of 0.2 ug/d for elemental 

analyses, 0.15 dpm/d for 234U 

and 238U and essentially 

anything detected for 235U 

(Bihl 2004, pg. 27) 

Solubility Class 

The default assumption of M or 

S would be appropriate, based 

on the most claimant-favorable 

result to the organ in question 

(TBD, pg. 10). 

For the maximizing approach, 

the most claimant-favorable 

solubility type for the organ of 

interest is used. For the best-fit 

approach the most appropriate 

solubility type can be used 

(Scalsky 2004, pg. 85). 

For the maximizing approach, 

the most claimant-favorable 

solubility type for the organ of 

interest is used. For the best-fit 

approach the most appropriate 

solubility type can be used. 

Inhalation class and lung 

absorption type for uranium is 

found in Bihl 2004, 

Table 5.2.5-3, pg. 24). 

Intake Date for 

Hypothetical Intake 

(excluding tritium) 

Not included in the TBD. 

Acute inhalation on January 1 

in the first year of employment 

(Scalsky 2004, pg. 85; Bracket 

2003, pg. 3). 

First day of employment or the 

first day of operation of the 

facility where the worker was 

assigned. For separation plants, 

chronic intakes would apply 

from either the first day of 

work for the worker or the 

start-up of the plant, December 

1944 for T Plant and April 

1945 for B Plant (Bihl 2004, 

pg. 8). 

Tritium Missed Dose 

Application 

Not included in the TBD. 

Assigned to workers monitored 

for external dose, but having 

no bioassay. For workers not 

in the dosimetry or bioassaymonitoring program, the 

missed internal dose is based 

on environmental intake only. 

Scalsky 2004, pg. 84; Duncan 

2003, pp. 6 and 12) 

Assigned to workers who 

worked in 108-B, the 300 Area 

Test Reactors, and in some 

cases where work location was 

unknown or variable. Those 

who never wore a dosimeter 

and had no bioassay results 

were assigned environmental 

doses (Bihl 2004, pp. 21–22). 
 

 

 

 

 

 

 

 

 

 

 

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Table A-3: Comparison of Default Assumptions for Internal Exposure at INL, 

Savannah River Site, and Hanford
 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Basis for Tritium 

Missed Dose Not included in the TBD. 

Dose calculated based on the 

tritium reporting level for a 

particular time period (Scalsky 

2004, pg. 67; Duncan 2003, 

Tritium urinalysis was not 

perfected until 1961. Liquid 

scintillation counting for 

tritium likely was started in 

1958 (Bihl 2004, pp. 21–22). 

From 1949 to 1960 the MDA 

was 5 uCi/L and from 1961 to 

1981 the MDA was 1 uCi/L. 

Later in 1982 the MDA 

changed to 10 dpm/ml and in 

1991 to 20 dpm/ml, (Bihl 2004, 

pg. 6). pg. 22). Tritium intakes were 

accounted for as part of 

external dose until about 1986– 

1987 (TBD does not explain 

methodology), when they were 

entered in the dose database as 

internal dose (Bihl 2004, pp. 12 

and 22). 

Hypothetical Intake 

Application Not included in the TBD. 

Applied to claims with nonmetabolic and digestive tract 

cancers (Scalsky 2004, pg. 85; 

Bracket 2003, pg. 2). 

Applied to individuals who 

wore a dosimeter but did not 

have any bioassay (Bihl 2004, 

pg. 48). 

Basis for missed 

internal dose from 

radionuclides other 

than tritium 

(1) If claimant file include 

positive external 

dosimeter readings, they 

should be treated as 

radiation workers and the 

default internal missed 

dose is applied as outlined 

in the table. If no 

detectable external or 

internal dose information 

in recorded, only the 

environmental dose 

should be included (TBD, 

pg. 37). 

(2) The probability that a 

worker received a 

significant unmonitored 

internal intake of 

radioactive material is 

very low. It is 

recommended that 

workers who have no 

recorded internal dose and 

wore a personnel 

dosimeter be treated the 

same as a worker who 

was monitored but had no 

bioassay results exceeding 

(1) Individuals with no 

external or internal 

monitoring data were 

assigned an 

environmental internal 

dose (Scalsky 2004, 

pg. 84; Bracket 2003, 

pg. 2). 

(2) For those individuals with 

external monitoring but 

no or limited internal 

monitoring, an annual 

missed tritium dose and 

environmental dose from 

uranium, plutonium, and

131I are assigned as 

internal dose. It is also 

reasonable to pick a 

fission or activation 

product that produces the 

largest dose to the organ 

of interest (Scalsky 2004, 

pg. 84; Bracket 2003, 

pg. 8). 

(3) Highest five intakes for 

various nuclides are 

applied to those 

individuals with non­

(1) Individuals with no 

external or internal 

monitoring data were 

assigned an environmental 

internal dose, (Bihl 2004, 

pg. 48) 

(2) For those individuals with 

external monitoring but no 

or limited internal 

monitoring, the approach 

was year dependent. For 

1947 through 1952, daily 

intakes at 10% of the 

respiratory protection 

required value for 40 

hours/week were assumed. 

Iodine was assumed to be 

at 0.1 times the vapor 

index. For 1953 through 

1988, daily intakes were 

based on an exposure to 

airborne concentrations at 

10% of the limiting air 

concentration for four 

hours per week, (Bihl 

2004, pg. 49). 

(3) From 1989 through the 

present, a daily exposure 
 

 

 

 

 

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Table A-3: Comparison of Default Assumptions for Internal Exposure at INL, 

Savannah River Site, and Hanford
 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

reporting levels. It is metabolic or digestive at 5% of the limiting air 

further recommended that system cancers (Bracket concentration for 4 hours 

individuals that were not 2003, pg. 2). per week was assumed, 

issued a personal (Bihl 2004, pg. 50). 

dosimeter and have no (4) For monitored workers 

record of internal dose with no confirmed intake, 

monitoring be assigned a maximum intake is 

only the environmental determined by using the 

dose for the facility MDA of the last sample as 

(TBD, pg. 39). an upper bound (Bihl 

(3) Construction workers that 2004, pg. 47). 

were issued personnel 

dosimeters should be 

treated the same as 

facility employees that 

were issued personnel 

dosimeters. Construction 

workers that were not 

issued a personnel 

dosimeter should be 

assigned the 

environmental dose for 

the facility (TBD, pg. 40). 

Variable by facility and organ 

of interest. Alpha intakes are 

assigned for the Plutonium 

Finishing Plant (PFP), the 200 

Area Fuel Separations Plants, 

U-Plant, C-Plant, the 300 Area 

Fuel Fabrication Facilities, 

209E, 120, 324, 325, 327, the 

Tank Farms and evaporator 

facilities (0.5 times the alpha 

intake), and where work 

Radionuclides 

included in the 

Hypothetical Intake 

Not included in the TBD. 

241Am/241Pu (M), 244Cm (M), 60Co (S), 137Cs (F), 237Np (M), 238Pu (M), 239Pu (M), 90Sr (F), 234U (F), and 238U (F) 

(Bracket 2003, pg. 9) 

location is unknown or highly 

variable. Alpha intakes are 

based primarily on 234U or 

239Pu. Beta/gamma intakes are 

assigned for all facilities except 

PFP, 209E, 120, the 300 Area 

Fuel Fabrication Facilities, 

108-B, and U-Plant. Tritium 

intakes are assigned for the 

108-B Building, the 300 Area 

Test Reactors, and in some 

situations where work locations 

are unknown or variable. 

The particular beta/gamma 

radionuclide and its solubility 

class are determined based on 

the organ of concern. For 
 

 

 

 

 

 

 

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Table A-3: Comparison of Default Assumptions for Internal Exposure at INL, 

Savannah River Site, and Hanford
 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

some facilities and periods of 

time it is specified (Bihl 2004, 

pp. 51–52). 

Default Activity 

Ratios Pu Mixture Not specified in the TBD. Ten-year old 12% plutonium 

mix (Scalsky 2004, pg. 66). Not specified in the TBD. 

Activity Fractions for 

other Mixtures 

Activity fractions are provided 

for ATR fuel gaseous releases 

in Table 5.6.6.1-1 (TBD, 

pg. 36). 

Activity fractions are provided 

flor SMC DU uranium isotopes 

in Table 5.6.1-1 (TBD, pg. 28). 

Activity fractions are facility 

dependent. The activity 

fractions are taken from the 

Internal Dosimetry Technical 

Basis Manual (WSRC 1990). 

The information for these 

ratios was obtained from safety 

analysis reports, personal 

interviews, open literature, etc. 

Activity fractions are provided 

for uranium mixtures, Table 

5.2.5-3, page 24, weapons and 

fuel grade plutonium, Table 

5.2.1-3 page 16, and recycled 

uranium impurities., 

Table5.2.5-2, page 24. Default 

mixtures based fission product 

urinalysis was developed by 

time period and organ of 

concern (Bihl 2004, pg. 10, 

Attachment D). 

Radionuclides of 

Concern for 

Monitored Workers 

Radionuclides of concern are 

identified for each of the eight 

INEEL facilities (TBD, Tables 

5.6.1-1 to 5.6.6.1-2, pp. 28– 

37). 

Radionuclides of concern were 

based on the in-vivo and invitro bioassay data of the 

individual (Scalsky 2004, 

pp. 66 and 67). Although the 

TBD provides activity fractions 

in Attachment A, it is not clear 

how these activity fractions are 

used in dose calculations. 

Radionuclides of concern were 

based on the in-vivo and invitro bioassay data of the 

individual, or the minimum 

detectable activity for a 

particular radionuclide. 

Radionuclide assumptions 

varied by facility and organ of 

interest (Bihl 2004, pg. 13). 

Tritium Dose for 

Monitored Workers 

Tritium is not identified as a 

key radionuclide of concern at 

INEEL facilities. 

Based on the reporting level if 

the tritium bioassay is less than 

this level, or the actual 

bioassay result if it is greater 

than the reporting level. 

Organically Bound Tritium and 

Tritium urinalysis was not 

perfected until 1961. Liquid 

scintillation counting for 

tritium likely was started in 

1958 (pp. 21–22). From 1949 

to 1960 the MDA was 5 uCi/L 

and from 1961 to 1981 the 

MDA as 1 uCi/L. Later in 

1982 the MDA changed to 10 

dpm/ml and in 1991 to 20 

dpm/ml (pg. 22). Tritium 

Stable Metal Tritides are not 

considered (Bracket 2003, 

pg. 6). 

intakes were accounted for as 

part of external dose until 

about 1986–1987 (TBD doses 

not explain methodology), 

when they were entered in the 

dose database as internal dose 

(pp. 12 and 22). (Bihl 2004, pp. 

12 and 22) 
 

 

 

 

 

 

 

 

 

 

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Table A-3: Comparison of Default Assumptions for Internal Exposure at INL, 

Savannah River Site, and Hanford
 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Internal Dose for 

radionuclides other 

than tritium 

See Table 5.7-1: defaults table 

for missed dose. This table 

provides the default 

assumptions for calculating 

missed dose of personnel 

working at a specific INEEL 

facility and during a specified 

time period (TBD, pp. 38–39). 

Based on either actual bioassay 

values or detection levels for 

bioassay techniques. For nonmetabolic cancers, the 

maximizing approach is used 

(Scalsky 2003, pg. 85). 

Based on either actual bioassay 

values for positive values. 

Based on a chronic intake over 

the entire exposure period with 

the last sample assumed to be 

at the MDA (Bihl 2004, 

pg. 47). 

Basis for prebioassay program 

doses 

Not included in the TBD. Not included in the TBD. Air concentration tolerance or 

limits, (Bihl 2004, pg, 7) 

Ingestion Not included in the TBD. Not included in the TBD. 

Assigned during periods were 

air sampling was used to 

determine internal dose. The 

quantity is based on the air 

concentration level or on the 

guidance provided in 

Estimation of Ingestion Intakes 

(NIOSH 2004). 

(Bihl 2004, pg. 8) 

Surrogate 

Radionuclide in 

IMBA for 65Zn/95Zr 

Not included in the TBD. 

137Cs used as a surrogate. 

Surrogate Adjustment factor = 

2.43. (Brackett 2003, pg. 9) 

Not included in the TBD. 

Surrogate 

Radionuclide in 

IMBA for 

106Ru/144Ce/95Nb 

Not included in the TBD. 

Radionuclides not available in 

IMBA. 90Sr used as a 

surrogate. Surrogate 

Adjustment factor = 7.25 

(Brackett 2003, pg. 9). 

Not included in the TBD. 

Surrogate 

Radionuclide in 

IMBA for 

242Cm/252Cf 

Not included in the TBD. 

Radionuclides not available in 

IMBA. 244Cm used as a 

surrogate. Surrogate 

Adjustment factor = 1.09 

(Brackett 2003, pg. 9). 

Not included in the TBD. 

IREP Radiation 

Types for 

Hypothetical Intake 

Not included in the TBD. 

Alpha 

Beta: >15 keV 

Tritium: < 15 keV 

(Bracket 2003, pp. 8 and 12) 

Alpha1 

Beta: >15 keV1 

Photon: > 250 keV1 

Tritium: < 15 keV1 

IREP Dose 

Distribution Type Not included in the TBD. Constant (Brackett 2003, pg. 12) Constant1 

Internal Dose 

Uncertainty Not included in the TBD. 

For the missed dose 

assignments, the value entered 

includes the uncertainty. 1

 No 

direction is provided to the 

dose reconstructor for dose 

assignments based on 

monitoring data. 

For the missed dose 

assignments, the value entered 

includes the uncertainty. For 

dose assignments based on 

monitoring data, the following 

values can be applied as a 

standard deviation: 

(1) 0.3 times the MDA or 

reporting level, or 

(2) 0.5 times the MDA for 

chest counting. 
 

 

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Table A-3: Comparison of Default Assumptions for Internal Exposure at INL, 

Savannah River Site, and Hanford
 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Actually report errors can be 

used if available (Bihl 2004, 

pg. 46). For air concentration 

data, a triangular distribution 

with zero as the minimum, the 

derived values as the mode, 

and twice the mode as the 

maximum is used (Bihl 2004, 

pg. 7). 

Other Comments 

Personnel employed in the 

Naval Program are not 

included in this TBD (TBD, 

pg. 7). However, through the 

years, NRF has participated in 

limited coordination of 

radiological protection 

programs and site support 

services. Some workers’ 

internal dose could have 

resulted from their support 

work at the NRF (TBD, pg. 9). 

None. 

Informs the dose reconstructor 

of limited use radionuclides 

such as 14C, 232Th, radon, 90Y, 227Th, 227Ac, and 32P (Bihl 

2004, pg. 32) 

1

 These parameters were obtained from review of several Hanford dose reconstruction IREP input sheets. 
 

 

 

 

 

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Table A-4: Comparison of Default Assumptions for Environmental Exposure at INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Application 

Environmental doses for 

personnel working at various 

INL facilities are calculated 

using facility-specific intake 

values and fence-line TLD 

direct gamma dose values 

(Tables 4-1 to 4-13, TBD). 

Apply the annual internal and 

external environmental dose 

for each full or partial year of 

employment for the 

maximizing approach. Dose 

reconstructors are instructed to 

use only the maximum annual 

intakes in Table C-17 for the 

maximizing approach (Scalsky 

2004, pg. 179). For the best-fit 

approach, modifications can be 

made for partial year of 

Environmental doses are 

assigned to personnel with no 

bioassay and no evidence of 

having worn a dosimeter at the 

Hanford Site 

(Bihl 2004, pg. 48). 

employment. No 

environmental dose is assigned 

if the background is not 

subtracted from the workers 

badge (Scalsky 2004, pg. 62). 

Sources of 

Environmental 

Releases Considered 

The TBD focuses mainly on 

airborne gaseous effluent 

releases and direct beta/gamma 

radiation from INL facilities. 

All releases considered in the 

Idaho National Engineering 

Laboratory Historical Dose 

Evaluation (DOE 1991a) are 

the basis for the releases 

considered in the TBD (TBD, 

pg. 8). Eight facility areas have 

been chosen for TBD analysis: 

TAN, ICPP, TRA, RWMC, 

CFA, SPERT, ARA, and 

ANL-W (TBD, pg. 10). 

The TBD heavily references 

the Cummins (1991) and CDC 

(2001) documents, and dose 

not include many of the base 

assumptions from those 

reports in the TBD. It is 

apparent that releases from the 

reactors and separations areas 

were considered. 

T-plant particles and iodine, 

B-Plant particles and iodine, 

REDOX particles and iodine, 

PUREX particles and iodine, 

Z-Plant particles, reactor noble 

gases, and tritium from 108B 

Building 

(Savignac 2003, pg. 18). 

Source Term Basis 

Idaho National Engineering 

Laboratory Historical Dose 

Evaluation Report (DOE 

1991[a]). Identification and 

Priorization of Radionuclide 

Releases from the Idaho 

National Engineering and 

Environmental Laboratory 

(RAC 2002). Version 6 of the 

RSAC code (Wenzel and 

Schrader 2001) is used 

extensively in the current 

report to provide onsite 

concentrations due to episodic 

releases as well as other 

evaluations (TBD, pg. 8). 

Radioactive Releases from the 

Savannah River Plant 1954– 

1989 (Cummins 1991), 

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), SRS meteorology data, 

SRS environmental reports for 

1993–2001. 

Hanford Works environmental 

reports; Methods for 

Estimating Radiation Doses 

from Short-Lived Gaseous 

Radionuclides and Radioactive 

Particles Released to the 

Atmosphere During Early 

Operations at Hanford (Till et 

al. 2002). 
 

 

 

 

 

 

 

 

 

 

Effective Date: 

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Table A-4: Comparison of Default Assumptions for Environmental Exposure at INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Methodology 

For routine operations, 

atmospheric dispersion of 

releases was modeled using 

annual average meteorological 

conditions. Dispersion of 

episodic releases was generally 

modeled using actual hourly 

wind speed and direction data 

at the time of release (BOE 

1991). Progressive Gaussian 

Gaussian model (Scalsky 

2004, Section 3.1.1) 

Puff advection (RATCHET) 

model (Savignac 2003, pg. 14) 

Type of Releases 

The TBD included two types 

of releases: (1) normal 

operational releases, and (2) 

episodic releases that generally 

are of short duration (e.g. 

criticality) (TBD, pg. 6). The 

TBD claims that these releases 

potentially represent 

unrecorded or missed does, 

either as direct gamma or betagamma from immersion in the 

radioactive gaseous cloud, for 

those individuals who do not 

have personal dosimetry to 

record the dose, or as internal 

doses from inhalation. 

The TBD heavily references 

the Cummins (1991) and CDC 

(2001) documents, and dose 

not include many of the base 

assumptions from those 

reports in the TBD. 

Calculations included routine 

and identified non-routine 

releases. Estimates include 

inhalation of radionuclides in 

air, direct external radiation 

from plumes, and physical 

contact with particulate 

radionuclides on skin. 

Ventilation Rate 

(m3/year) Not included in the TBD. 

2,400 (default); Adjustments 

can be made for light and 

heavy work (Scalsky 2004, 

pg. 162). 

2,400 (default); Based on 

1.2 m3

/hour ± 0.4 m3/hour 

(Savignac 2003, pg. 16) 

Exposure Time 

(hours/week) Not included in the TBD. 

40 with a 1.25 conversion 

factor to increase the exposure 

time to 50 hours/week 

(Scalsky 2004, pg. 61). 

40 (Savignac 2003 pg. 24) 

Mobile Workforce Information not included in the 

TBD. 

Assign the maximum dose 

listed for any area onsite. 

Information not included in the 

TBD. 

Facility Specific 

Workforce 

Intake values are listed for 

eight specific facilities 

including TAN, ICPP, TRA, 

RWMC, CFA, SPERT, ARA, 

and ANL-W (TBD, pg. 10). 

Assign the maximum dose 

listed for any area onsite for 

the maximizing approach. 

Assign an area specific 

environmental dose based on 

the work location of the 

worker for the best-fit 

approach (Scalsky 2004, 

pg. 61). 

Information not included in the 

TBD. 

Radionuclides 

Considered for 

External Dose 

No specific radionuclides 

considered in the TBD. Direct 

gamma values were taken 

from fence-line TLD reading 

(TBD Table 4-13). 

41Ar, (Scalsky 2004, pg. 60) 

41Ar,131I, 106Ru (Savignac 

2003, pp. 19 and 23) 
 

 

 

 

 

 

 

 

Effective Date: 

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Table A-4: Comparison of Default Assumptions for Environmental Exposure at INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

Radionuclides 

Considered for 

Submersion Dose 

Noble gas and halogen (TBD, 

pg. 24). 

41Ar, (Scalsky 2004, pg. 59) 

41Ar, page 17, 131I, 3

H 

Kathy – can’t find evidence 

that these last two belong here. 

Submersion DCF 

Not considered. TBD states 

that air immersion doses 

would be recorded in the 

fence-line TLD doses 

presented in Table 4-13. 

Assumed values from the 

Federal Guidance Report 12 

(EPA 1993). 

(Scalsky 2004, pg. 60) 

Federal Guidance Report No. 

13, Cancer Risk Coefficients 

for Environmental Exposure to 

Radionuclides, 1999. 

Radionuclides 

Considered for 

Internal Dose. 

Facility Annual Intake (Bq/yr) 

due to normal operations 

provided for Ce-144, I-131, 

Pm-147, Pu-238, Pu-239/240, 

Ru-106, Sr-89, Sr-90, and 

Y-91 (TBD Tables 4-1 to 4-8). 

Specific Incident Intake 

(Bq/event) due to criticality 

provided for Rb-89, Sr-91, Sr92, Y-92, Y-93, Te-133, I-131, 

I-133, I-134, I-135, Cs-138m 

Ba-139, La-141, La-142, and 

U-234 (TBD Table 4-10). 

Individual Test Intake 

(Bq/event) due to Special 

Tests provided for Sr-89, Sr90, Y-91, Zr-95, Ru-103, Ru106, I-131, Ce-144, and 

Pr-143 (TBD Table 4-11). 

Individual Test Intake 

(Bq/event) due to Initial 

Engine Tests provided for 

Rb-89,Sr-89, I-131, I-133, 

I-135, Cs-138, and U-234 

(TBD Table 4-12). 

3

H,131I, 238Pu, 239Pu, 240Pu, 234U, 235U, and 238U 

(Scalsky 2004, pg. 51) 

3

H,131I-131mXe, 144Ce-144Pr, 137Cs-137Ba, 239Pu, 103Ru­ 103mRh, 106Ru-106Rh, 90Sr-90Y,

95Zr-95Nb 

(Savignac 2003, pg. 8) 

Soil Not included in the TBD. 

Density = 1,600 kg/m3 

Surface Factor = 0.08 

Resuspension Factor =1E-9/m 

(Scalsky 2004, pg. 59) 

Not included in the TBD. 

Liquid Effluents Not included in the TBD. Not included in the TBD. Not included in the TBD 

Organ Dose 

Conversion Factor Not included in the TBD. 

1.0 is used in the maximizing 

approach. The organ dose 

conversion factors in the 

external dosimetry guide for 

the relevant exposure 

geometry are used in the bestfit analysis 

(Scalsky 2004, pg. 61). 

Not included in the TBD. 

IREP Rate Not included in the TBD. Chronic (Scalsky 2004, pg. 61) Chronic1 
 

 

 

 

 

 

Effective Date: 

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Table A-4: Comparison of Default Assumptions for Environmental Exposure at INL, 

Savannah River Site, and Hanford 

Description of 

Assumption Idaho National Laboratory SRS Hanford 

IREP Radiation 

Type Not included in the TBD. 

Photon, 30–250 keV

41Ar , 100% photon, > 250 

keV (Scalsky 2004, pp. 60 and 

61) 

Photon, 30–250 keV1 

IREP Dose 

Distribution Type |Not included in the TBD. 

Constant. Doses and intake 

quantities provided with a 50th 

percentile and a geometric 

standard deviation. A 95th 

percentile for the source term 

is estimated as 25% greater 

than the 50th percentile 

(Scalsky 2004, pg. 60). 

Constant. Doses and intake 

quantities provided with a 

geometric mean and standard 

deviation. There is no 

direction on how these values 

should be entered into IREP. 

Special 

Considerations for 

Uranium and 

Plutonium 

No special considerations for 

uranium and plutonium in the 

TBD. 

The isotope yielding the 

maximum organ dose was 

assumed at 100% rather than 

applying a mixture (Scalsky 

2004, pg. 59). 

Not applicable. 

Other 

Dose reconstruction for 

individual whose location is 

unknown should use intakes 

provided by Table 4-3 (CFA) 

and exposures for ICPP as 

provided in Table 4-13. The 

values suggested will 

maximize the resultant 

individual dose. 

1955 values are assigned to 

1952, 1953, and 1954 (Scalsky 

2004, pg 54) 

The four chemical separations 

plants, T Plant, B Plant, 

REDOX Plant and the PUREX 

plant, along with the 

plutonium handling Z-plant are 

shown in Figure 4.1.1 to be the 

most important release points 

at Hanford (Savignac 2003). 

1

 These parameters were obtained from review of several Hanford dose reconstruction IREP input sheets. 
2 

Effective Date: 

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ATTACHMENT 5: ISSUE RESOLUTION MATRIX FOR FINDINGS AND KEY OBS

Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment Finding SC&A TBD Number Issue Number and Description NIOSH Response 

Number Number Page No. 

ORAUT­ 1 5 TKBS-0007-4 

Issue 1: (5.1.1.1) Routine Airborne Releases - Source terms provided require improvement for use in determining the worker intake from airborne releases at different INL facilities. The data NIOSH uses do not take into account the deficiencies in the environmental monitoring equipment and their locations, and, in addition, NIOSH does not assess the uncertainties associated with the meteorological dispersion model used for the INL site. Most importantly, the source terms do not account for worker inhalation of resuspended contaminated soils and materials around the INL facilities. 

45

ORAUTTKBS-0007-4 

Issue 2: (5.1.1.2) Episodic Airborne Release - The airborne releases associated with several of the Initial Engine Tests of the Aircraft Nuclear Propulsion (ANP) Program were likely to have 6 55 been underestimated by factors ranging from 2 to 7. Also, NIOSH did not evaluate the uncertainties associated with the deficiencies in air monitoring equipment. 

3 ORAUT­ TKBS-0007-4 7 

Issue 3: (5.1.1.3) Direct Gamma Exposures – The fence-line TLD measurements are not adequate for reconstructing direct gamma doses to personnel working outdoors at and around a specific INL facility inside the fence-line boundary, because they do not take into account the most bounding scenarios. 

57 

4 ORAUT­ TKBS-0007-5 8 Issue 4: (5.1.2.1) Completeness and Quality of INL Internal Dosimetry Programs - The identification and determination of missed internal 73 
Effective Date: 

January 25, 2006 

Revision No.  1 (Draft) 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

dose for workers are heavily influenced by the assumption of confidence, but SC&A found this premise to be unsupported after examining several critical DOE-HQ Tiger Team and DNFSB site audit reports. In addition, many site experts interviewed by SC&A indicated that there were significant deficiencies and inconsistencies in radiation work practices throughout the operating history of the INL facilities. These observations jeopardize the validity of the TBD approaches in reconstructing missed worker internal doses. 

5 ORAUT­ TKBS-0007-5 9 

Issue 5: (5.1.2.2) High-Risk Jobs (Internal Exposure) - NIOSH did not evaluate comprehensively the facility and field data to identify and separate out the high-risk or high-dose jobs for worker internal exposures. This information is essential for dose reconstructors to 

77 

fill in the data gap when dose records in a claimant’s file are not complete. 

6 ORAUT­ TKBS-0007-5 O 

Issue 6: (5.1.2.3) Calibration of Internal Dosimetry Analytical and Monitoring Equipment - The TBD does not provide any information on the calibration procedures, sensitivities, and standards of the internal dosimetry analytical equipment and monitoring instrumentation. The 1991 DOE Tiger Team findings show the deficiencies in these areas. NIOSH should evaluate the uncertainties 

78 

and impacts on the internal dose assessment results associated with the deficient calibration programs at INL. 

7 ORAUT­ TKBS-0007-5 O 

Issue 7: (5.1.2.4) Changes of Internal Dose Limits - Inconsistent work practices were prevalent in the early years of the INL operation and may have led to significant missed dose to workers. NIOSH should evaluate the impacts of these dose limit 

78 
Effective Date: 

January 25, 2006 

Revision No.  1 (Draft) 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

changes over the operating history of INL to see whether there were missed doses in the early years when the radiation protection policy was less protective and inconsistently implemented. 

8 9 

ORAUTTKBS-0007-5 ORAUTTKBS-0007-5 

10 

O 

Issue 8: (5.1.2.5) High-Fired Plutonium and Uranium Intakes - The TBD did not evaluate the hazard associated with high-fired plutonium and uranium at the INTEC (ICPP) and RWMC facilities. High-fired Pu-238, Pu-239, and uranium are not easily dissolvable, nor do they readily break into very small particles. They also emit some gamma rays and neutrons. Similar to the treatment of recycled uranium, NIOSH should evaluate the lung dose for intake of high-fired uranium and plutonium oxide particulates (alveolar deposition). Issue 9: (5.1.2.6) Skin and Facial Contamination - This TBD does not consider incidents with workers having skin contamination, facial contamination, and positive nasal swipes in the INL facilities. These kinds of problems would be compounded by the deficiencies in air sampling systems and ineffective respiratory protection programs. Guidance should be provided to a dose reconstructor to account for the missed dose due to the unaccounted uptake. 

78 79 

10 11 

ORAUTTKBS-0007-5 ORAUTTKBS-0007-5 

O 

11 

Issue 10: (5.1.2.7) Breathing Rates - The TBD assumption appears less claimant favorable than the ICRP or NCRP assumptions. Issue 11: (5.1.2.8) Non-Occupational Worker Elimination of DU Background - The derivation of the background value of 0.16 μg/L used for subtraction from each urinalysis result of uranium prior to assessment of occupational internal dose for SMC radiation workers is not technically 

79 79 
14 

Effective Date: 

January 25, 2006 

Revision No.  1 (Draft) 

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Page No. 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

sound. The baseline background (population) intake value was determined by a study of urine samples submitted by non-radiation workers at the SMC facility. A better approach would be to use the urine excretion samples by non-INL people in the Idaho Falls areas. NIOSH should consider this subtraction from urinalysis results as a missed internal dose. 

12 ORAUT­ TKBS-0007-5 O 

Issue 12: (5.1.2.9) Unmonitored Workers - The potential missed doses for unmonitored workers would be from inhaling resuspended contaminated soils and ingesting contaminated materials while eating in a contaminated, previously considered uncontaminated, area (such as office and cafeteria). NIOSH should evaluate these potential missed doses. 

80 

13 

ORAUTTKBS-0007­

4/5/6 

O 

Issue 13: (5.1.2.10) Naval Reactor Facility Workers - As the internal dose TBD indicates, “some workers’ internal dose could have resulted from their support work at the NRF.” NIOSH should evaluate the potential missed dose at the NRF for these workers. 

80 

ORAUTTKBS-0007-5 

Issue 14: (5.1.2.11) Plutonium Monitoring - The TBD does not provide any historical information on the plutonium analysis methods used at INL. It is entirely possible that selective plutonium monitoring on workers was used at INL until O 80 1980, but without this information, the dose reconstructors would not be able to assign missed internal dose due to plutonium intakes in the time period before 1980. NIOSH should provide information on plutonium monitoring. 

ORAUT- Issue 15: (5.1.3) SL-1 Accident Dose 

15 TKBS-0007­ 1 Reconstructions - The TBDs do not evaluate the 80 

4/5/6 potential missed internal and external doses or the 
Effective Date: 

January 25, 2006 

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Page No. 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment Finding SC&A TBD Number Issue Number and Description NIOSH Response 

Number Number Page No. 

associated uncertainties for the over 1,000 rescue and cleanup workers involved with the SL-1 accident that occurred in January 1961. There was a high potential for significant exposures, because the equipment used and the radiological control policies in place in that era were not as advanced and protective as those in current use. The TBDs should develop adjustment factors related to stay time, dose field estimates, internal dose results, external dose readings, and contamination level estimates. 

16 ORAUT­ TKBS-0007-6 8 

Issue 16: (5.1.4.1.1) Completeness and Quality of INL Beta/Gamma Dosimetry and Record Keeping Programs - The identification and determination of missed external dose for workers are heavily influenced by this assumption of confidence, but SC&A found this premise to be unsupported after examining several critical DOE-HQ Tiger Team and DNFSB site audit reports. In addition, many site experts interviewed by SC&A indicated that there were significant deficiencies and inconsistencies in radiation work practices throughout the operating history of the INL facilities. These observations jeopardize the validity of the TBD approaches in reconstructing missed worker external doses. 

96 

17 ORAUT­ TKBS-0007-6 4 

Issue 17: (5.1.4.1.2) Penetrating and NonPenetrating Doses - NIOSH should re-evaluate the missed gamma dose, due to the deficiencies in the procedures and algorithms. 

96 

18 ORAUT­ TKBS-0007-6 O 

Issue 18: (5.1.4.1.3) Correction For Beta Doses – NIOSH should develop a method to consistently account for uncertainties in dosimetry readings. Claimant-favorable correction factors should be 

97 

developed for beta dose reconstruction. 
Effective Date: 

January 25, 2006 

Revision No.  1 (Draft) 

Document No. 

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Page No. 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

Issue 19: (5.1.4.1.4) Angular Dependence Correction Factor for Gamma Dose - NIOSH should provide angular dependence (anatomic geometry) correction factors for external gamma 

19 ORAUT­ TKBS-0007-6 O doses, particularly for low-photon energies, where the angular dependence of the sensitivity of the dosimeter is most pronounced. These correction 99 

factors are used to account for, for example, the bias introduced by a dosimeter worn at the neck level and the higher doses received by tissues/organs below the waist. Issue 20: (5.1.4.1.5) Restating Beta Dose As 

20 ORAUT­ TKBS-0007-6 O Gamma Dose - It is not claimant favorable to state that the entire dose measured in the open window 99 

is due to the beta dose. Issue 21: (5.1.4.1.6) Photon Spectrum Split – 

21 ORAUT­ TKBS-0007-6 O NIOSH should provide guidance assigning dose values for the 30 keV<E<250 keV and E>250 keV 99 

regions. Issue 22: (5.1.4.1.7) Immersion Dose - The dose recorded on a dosimeter due to a semi-infinite 

22 ORAUT­ TKBS-0007-6 O cloud irradiation would be approximately half of the actual dose received. NIOSH should, 100 

therefore, consider a weighting factor of 2 for immersion dose. Issue 23: (5.1.4.1.8) High-Risk Jobs (Beta/Gamma Exposure) - Site experts interviewed by SC&A classified INL as an “acute dose” site, with a significant number of facilities, operations, 

23 ORAUT­ TKBS-0007-6 9 experiments, and occurrences providing the possibility of personnel receiving dangerous levels 100 

of radiation. NIOSH did not evaluate comprehensively the facility and field data to identify and separate out the high-risk or high-dose jobs for worker external exposures. This 
Effective Date: 

January 25, 2006 

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Document No. 

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Page No. 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

information is essential for dose reconstructors to fill in the data gap when dose records in a claimant’s file are not complete. 

24 25 

ORAUTTKBS-0007-6 ORAUTTKBS-0007-6 

O O 

Issue 24: (5.1.4.1.9) Extremity Dose - NISOH should evaluate the potential for missed extremity dose for workers working in facilities where highly contaminated equipment, piping, instruments, valves, and systems resulted in exposures in confined spaces to hands. Issue 25: (5.1.4.1.10) Discrepancies between PIC and Film Reading – NIOSH should compare PIC versus film badge data (i.e., shallow and deep), and ensure that all the dose has been captured by the film badge. It is important to note that some PICs were worn for only the length of the job, so the discrepancy between readings of the twodosimeter systems cannot be explained by drifting. 

100 100 

26 27 28 

ORAUTTKBS-0007-6 ORAUTTKBS-0007-6 ORAUTTKBS-0007-6 

O 

3 3 

Issue 26: (5.1.4.1.11) Minimum Detection Limit – NIOSH should re-evaluate the approach in determining the MDL of the dosimetry system by taking into account the system uncertainties. Issue 27: (5.1.4.1.12) Minimum Reporting Level (Beta/Gamma) - NIOSH does not provide adequate information supporting the use of chosen detection threshold levels to represent the MRL values for gamma film badges and TLDs. The use of MRL/2 as the missed external dose for dose reconstruction per OCAS-IG-001 is not claimant favorable for claims where the probability of causation value is close to 50%. In addition, NIOSH should re-evaluate the MRL values used and provide more supportable default values. Issue 28: (5.1.4.2.1) Minimum Reporting Level (Neutron) - NIOSH’s approach for determining the MRL values for NTA emulsion film is not 

101 103 108 
Effective Date: 

January 25, 2006 

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Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

thorough or supported. For example, NIOSH uses 10 neutron readings in one data sheet from March 1958 to determine the MRL values for the period between 1951 and 1957, and 6 neutron readings to represent all neutron measurements between 1959 and 1976. Furthermore, the use of MRL/2 as the missed external dose for dose reconstruction per OCAS-IG-001 is not claimant favorable for claims where the probability of causation value is close to 50%. In addition, NIOSH’s MRL values of 14 mrem and 20 mrem appear low and are inconsistent with generic values given for NTA dosimeters, as well as values cited by other DOE facilities with similar neutron source terms and detectors. NIOSH should re-evaluate the MRL values used and provide more supportable default values. 

29 ORAUT­ TKBS-0007-6 2 

Issue 29: (5.1.4.2.2) Failure to Properly Address Neutron Exposures - INL had a total of 52 reactors, most of which were experimental/ prototype in design, which typically operated with high-power densities and with minimum shielding and neutron moderation. It is unjustified to presume that there are no missed neutron doses. In addition, there are deficiencies associated with 109 neutron calibrations. Due to the use of the PoBe source for neutron calibration, dosimeters would significantly under-measure neutron doses from sources with lower-energy spectra. NIOSH should re-evaluate the entire approach in the TBD to account for potential missed neutron doses. 

30 ORAUT­ TKBS-0007-6 2 

Issue 30: (5.1.4.2.3) Neutron Calibration Deficiencies - Due to the use of the PoBe source for neutron calibration, dosimeters would 110 significantly under-measure neutron doses from 
Effective Date: 

January 25, 2006 

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Page No. 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

sources with lower energy spectra. NIOSH should re-evaluate the approach in the TBD to account for potential missed neutron doses. 

31 32 33 

ORAUTTKBS-0007-6 ORAUTTKBS-0007-6 ORAUTTKBS-0007-6 

8 

O O 

Issue 31: (5.1.4.2.4) Completeness and Quality of INL Neutron Dosimetry and Record Keeping Programs - The identification and determination of missed neutron dose for workers are heavily influenced by this assumption of confidence, but SC&A found this premise to be unsupported after examining several critical DOE-HQ Tiger Team and DNFSB site audit reports. In addition, many site experts interviewed by SC&A indicated that there were significant deficiencies and inconsistencies in radiation work practices throughout the operating history of the INL facilities. These observations jeopardize the validity of the TBD approaches in reconstructing missed worker neutron doses. Issue 32: (5.1.4.2.5) Uncertainty Estimation for Neutron Doses – NIOSH should explain how the FNCFs were obtained and provide instruction to dose reconstructors on how to apply them. Issue 33: (5.1.4.2.6) Neutron Organ Dose – NIOSH should provide neutron spectrum information and guidance for organ dose reconstruction for workers at ZPPR and TREAT. 

110 110 110 

34 ORAUT­ TKBS-0007-6 9 

Issue 34: (5.1.4.2.7) High-Risk Jobs (Neutron Exposure) - NIOSH did not evaluate comprehensively the facility and field data to identify and separate out the high-risk or high-dose jobs for worker neutron exposures. This information is essential for dose reconstructors to fill in the data gap when dose records in a claimant’s file are not complete. 

111 
Effective Date: 

January 25, 2006 

Revision No.  1 (Draft) 

Document No. 

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Page No. 

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Table A-5: Issue Resolution Matrix for INL Findings and Key Observations 

Comment 

Number TBD Number Finding Number Issue Number and Description SC&A Page No. NIOSH Response 

Issue 35: (5.1.4.2.8) Multiplying Factors for 

35 ORAUT­ TKBS-0007-6 O Missed Neutron Dose – NIOSH should provide data to support the two multiplying factors (1.25 and 2) and the fixed missed neutron dose default 111 

value of 50 mrem. 

Note: O-Observation