SEC Petition Evaluation Report

Petition SEC-00219 

Report Rev #:0 Report Submittal Date: March 12, 2015

Subject Expert(s): Mitch Findley, Brian Gleckler, Jason Davis, Mike Mahathy

Site Expert(s): N/A

Petition Administrative Summary

Petition Under Evaluation

Petition # Petition 

Type

Petition

Receipt Date

Qualification Date DOE/AWE Facility Name

SEC-00219 83.13 July 8, 2014 September 16, 2014 Idaho National Laboratory

Petitioner-Requested Class Definition 

All employees who worked in any area of the Idaho National Laboratory from January 1, 1949 through December 

31, 1970.

Class Evaluated by NIOSH

All employees who worked in any area of the Idaho National Laboratory from January 1, 1949 through December 

31, 1970.

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

All employees of the Department of Energy, its predecessor agencies, and their contractors and subcontractors

who worked at the Idaho National Laboratory in Scoville, Idaho, and were monitored for external radiation at the 

Idaho Chemical Processing Plant (CPP) (e.g., at least one film badge or TLD dosimeter from CPP) between 

January 1, 1963 and December 31, 1974 for a number of work days aggregating at least 250 work days, occurring 

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

or more other classes of employees in the Special Exposure Cohort.

Related Petition Summary Information

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

N/A N/A N/A N/A

Related Evaluation Report Information

Report Title DOE/AWE Facility Name

N/A N/A

ORAU Lead Technical Evaluator: Mitch Findley ORAU Peer Review Completed By: Dan Stempfley

Peer Review Completed By: [Signature on File]

Timothy Taulbee

3/12/2015 

Date

SEC Petition Evaluation Reviewed By: [Signature on File]

James W. Neton

3/12/2015 

Date

SEC Evaluation Approved By: [Signature on File]

Stuart L. Hinnefeld

3/12/2015 

Date

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Evaluation Report Summary: SEC-00219, Idaho National Laboratory

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

addresses a class of employees proposed for addition to the Special Exposure Cohort (SEC) per the

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended, 42 U.S.C. 

§ 7384 et seq. (EEOICPA) and 42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. 

Petitioner-Requested Class Definition

Petition SEC-00219 was received on July 8, 2014, and qualified on September 16, 2014. The 

petitioner requested that NIOSH consider the following class: All employees who worked in any area 

of the Idaho National Laboratory from January 1, 1949 through December 31, 1970. 

Class Evaluated by NIOSH

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

the following class: All employees who worked in any area of the Idaho National Laboratory from 

January 1, 1949 through December 31, 1970. 

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

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

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes all employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Idaho National Laboratory in Scoville, Idaho, and 

were monitored for external radiation at the Idaho Chemical Processing Plant (CPP) (e.g., at least one 

film badge or TLD dosimeter from CPP) between January 1, 1963 and December 31, 1974 for a 

number of work days aggregating at least 250 work days, occurring either solely under this 

employment, or in combination with work days within the parameters established for one or more 

other classes of employees in the Special Exposure Cohort. 

The proposed class differs from the class selected for evaluation by NIOSH because the operating area 

and time period for which dose reconstruction is deemed infeasible for the period evaluated in this 

report is limited to the CPP. The SEC class start date was established as January 1, 1963 because, in 

the face of increased alpha contamination, the CPP radiation protection program did not initiate a 

corresponding increase in the number of bioassay samples taken. The Preliminary ICPP Health 

Physics Upgrade Program report was published and delivered to management in October 1974. 

Although the proposed class end date is identified as December 31, 1974, further NIOSH-ORAUT 

review and assessment will be necessary to determine if the program data from 1975 forward is 

adequate to support reconstructing dose with sufficient accuracy for this area.
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Feasibility of Dose Reconstruction

NIOSH finds it is not feasible to estimate internal exposures with sufficient accuracy for all workers 

who worked in the Chemical Processing Plant at INL from January 1, 1963 through December 31, 

1974. With the exception of this class, per EEOICPA and 42 C.F.R. § 83.13(c)(1), NIOSH has 

established that it has access to sufficient information to: (1) estimate the maximum radiation dose, for 

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

plausible circumstances by any member of the class; or (2) estimate radiation doses more precisely 

than an estimate of maximum dose. Information available from the site profile and additional 

resources is sufficient to document or estimate the maximum internal and external potential exposure 

to members of the evaluated class under plausible circumstances during the specified period. For the 

remainder of the INL site, other than the locations and periods defined as reserved for further 

assessment and evaluation, dose reconstructions are deemed feasible at INL from January 1, 1949 

through December 31, 1970. 

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

• Operations at the site began in 1949, when the Atomic Energy Commission (AEC) established the 

National Reactor Testing Station (NRTS). The site, known today as the Idaho National 

Laboratory (INL), was the primary nuclear reactor development laboratory in the United States. 

Over 100 reactor concepts were conceived and tested there. Between 1953 and 1992, the Idaho 

Chemical Processing Plant (CPP) at INL reprocessed spent nuclear fuel from naval propulsion, 

test, and research reactors to recover enriched uranium for reuse in nuclear weapons production. 

Other facilities at INL also conducted various nuclear weapons research and development 

activities over the INL operational period evaluated in this report.

• Principal sources of internal radiation for members of the proposed class included inhalation and 

ingestion exposures to various isotopes of uranium, thorium, and plutonium, tritium, exotic 

radionuclides (produced from, or as a result of, reactor neutron irradiation), and mixed fission and 

activation products (MFP/MAP). These materials were most widely used and present on site at 

INL in the Test Reactor Area (TRA), Chemical Processing Plant (CPP), Test Area North (TAN), 

Miscellaneous Reactor Areas, Central Facilities Area (CFA), and Burial Ground. 

o TRA: NIOSH has determined that the internal dose potential during the period under 

evaluation at the TRA is associated with reactors, laboratories, machine shops, and other 

research and support activities as well as radioactive waste disposal. There was limited 

personnel internal exposure potential to alpha-emitting radionuclides, including plutonium and 

other actinides, during the evaluated period. Personnel monitoring data for mixed fission 

products exposures exist to support reconstructing doses for INL workers at during the 

evaluated period. 
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o CPP: NIOSH has determined that the internal dose potential at CPP over the evaluated period 

was related to its primary function of processing spent fuel elements containing enriched 

uranium for the recovery of the un-fissioned uranium. The uranium was separated from 

fission products by continuous liquid-liquid extraction. Although much of the processing 

equipment was heavily-shielded and operated remotely, the plant design was based on a 

direct-contact maintenance philosophy. There was known alpha contamination (from 

unspecified alpha-emitting radionuclides) in the Analytical Labs, a routine exposure potential 

from the Np-Pu campaign, routine exposure potential from the Umpire Program, radiological 

program degradation including decreased personnel monitoring, and limited plutonium 

monitoring data over the evaluation period. 

o TAN: NIOSH has determined that the internal dose potential at TAN was related to the initial 

work for the Aircraft Nuclear Propulsion (ANP) Program, which was eventually terminated, 

after which operations shifted to the Safety Test Engineering Program (STEP). The overall 

objectives at TAN included conducting full-scale tests on reactor systems to determine what 

actually happens to fission products released during a reactor accident, demonstrating reactor 

safety, developing realistic analytical methods useful for future reactor designs, and using

basic information already developed in the Nuclear Safety Program. There was no appreciable 

exposure to actinides without a correlating exposure to MFPs, which allows for relating the 

MFP and actinide exposure data.

o Misc. Reactor Areas: NIOSH has determined that internal exposures in the Miscellaneous 

Reactor Areas, which includes the Auxiliary Reactor Area (ARA), Organic Moderated Reactor 

Experiment (OMRE), and the Special Power Excursion Reactor Tests (SPERT), are related to 

reactor and hot cell operations and testing. The exposures in these areas are primarily related 

to MFP for which personnel monitoring data are available. One exception involved the 

separation of Pa-233 from irradiated thorium slugs at the ARA-I hot cells in 1968. NIOSH 

will be evaluating this work and related personnel monitoring separately. 

o CFA: NIOSH has determined that internal exposures at the CFA were related to the laundry 

facility, support laboratory, and support machine shop operations. The exposures in these 

areas are primarily related to uranium and MFP for which personnel monitoring data are

available.

o Burial Ground: NIOSH has determined that internal exposures at the Burial Ground were 

directly related to the materials being disposed of in the grounds. Up to the point in time that 

drum retrieval commenced in 1969, exposure potential was virtually all from mixed fission

products in the INL waste being buried, and plutonium for the Rocky Flats Plant waste that 

was received for disposal. Internal monitoring data are available for the workers who 

supported the waste disposal activities. While data are also available for the drum-retrieval 

work that started in 1969, the exposure potential and issues are necessarily different and 

requires additional assessment and analysis. 
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• NIOSH has determined that it has access to sufficient personnel and area monitoring data to 

support internal dose reconstruction with sufficient accuracy for INL workers during the time 

period evaluated in this report, with the exception of workers at CPP from January 1, 1963 

through December 31, 1974 and reserved portions of the class, which include: 

o TAN: TAN-607 (Fuel Storage Vaults) and TAN-615 (Actuator Building) from 1961-1970 due 

to potential uranium exposures without mixed fission products being present 

o Misc. Reactor Areas: ARA-I in 1968 due to potential unmonitored exposure to Pa-233

o Burial Ground: The years 1969 forward due to a newly-implemented procedure of waste 

exhumation and retrieval 

• For the period from January 1, 1963 through December 31, 1974 at CPP, NIOSH has not located 

sufficient personnel or area monitoring documentation to support complete reconstruction of 

internal personnel exposures to uranium, neptunium, plutonium, and other related transuranic 

radionuclides. Without additional personnel radiation monitoring data or air monitoring data

during this period, NIOSH has insufficient information to appropriately characterize radioactive 

material intakes of these radionuclides during these INL operations. 

• Principal sources of external radiation for members of the proposed class included beta-gamma, 

and neutron exposures from various isotopes of uranium, thorium, and plutonium, tritium, exotic 

radionuclides (produced from, or as a result of, reactor neutron irradiation), and mixed fission and 

activation products (MFP/MAP). These materials were most widely used and present on site at 

INL in the Test Reactor Area (TRA), Chemical Processing Plant (CPP), Test Area North (TAN), 

Miscellaneous Reactor Areas, Central Facilities Area (CFA), and Burial Ground. By policy, INL

monitored by dosimeter all personnel who were expected to receive any radiation dose or whose 

work was centered at the site. The six major areas below required external dosimetry (film or 

TLD) be worn within the operating (fenced) area; this dosimetry was issued at the guard 

checkpoint; when a worker went from one area to another, a dosimeter was issued at each location.

o TRA: NIOSH has determined that the beta-gamma external dose potential during the period 

under evaluation at the TRA (based primarily on MFP) is associated with reactors, 

laboratories, machine shops, and other research and support activities as well as radioactive 

waste disposal. The neutron external dose potential during the period under evaluation at the 

TRA is associated with reactors and sources, and also from the Fast Chopper and Crystal 

Spectrometer. 

o CPP: NIOSH has determined that the beta-gamma external dose potential during the period 

under evaluation at the CPP is associated with fuel reprocessing, handling raffinates and 

calcined wastes, handling fuel cladding, and Radioactive Lanthanum (RaLa) work. The 

neutron external dose potential during the period under evaluation at the CPP was associated 

primarily with the handling of transuranic (TRU) isotopes and spontaneous fission of 

radionuclides. 
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o TAN: NIOSH has determined that the beta-gamma external dose potential during the period 

under evaluation at the TAN (based primarily on MFP) is associated with fuel nuclear reactors, 

activated materials, and irradiated reactor fuels. The neutron external dose potential during the 

period under evaluation at the TAN was associated primarily with operation of nuclear 

reactors and exposure to irradiated reactor fuels. 

o Misc. Reactor Areas: NIOSH has determined that the beta-gamma external dose potential 

during the period under evaluation at the Misc. Reactor Areas (based primarily on MFP), 

which includes the Auxiliary Reactor Area (ARA), Organic Moderated Reactor Experiment 

(OMRE), and the Special Power Excursion Reactor Tests (SPERT) is associated with reactors, 

laboratories, and other research and support activities. The neutron external dose potential 

during the period under evaluation at the Misc. Reactor Areas is associated with reactors and 

sources.

o CFA: NIOSH has determined that the beta-gamma external dose potential during the period 

under evaluation at the CFA (based primarily on MFP) is associated with residual exposure to 

radionuclides from the other facilities. The neutron external dose potential during the period 

under evaluation at the CFA is unlikely, but would be associated with residual exposure to 

radionuclides from the other facilities if it did exist.

o Burial Ground: NIOSH has determined that the beta-gamma external dose potential during the 

period under evaluation at the Burial Ground (based primarily on MFP) is associated with 

residual exposure to radionuclides based on disposal of wastes from the other facilities. The 

neutron external dose potential during the period under evaluation at the Burial Ground is 

unlikely, but would be associated with residual exposure to radionuclides in the wastes from 

the other facilities if it did exist.

• NIOSH has demonstrated that it has access to sufficient personnel and area monitoring data to 

support external dose reconstruction with sufficient accuracy for the worker class and time period 

evaluated in this report. 

• NIOSH finds that it is likely feasible to reconstruct occupational medical dose for INL workers 

with sufficient accuracy for the worker class and time period evaluated in this report.

Pursuant to 42 C.F.R. § 83.13(c)(1), NIOSH determined that there is insufficient information to either: 

(1) estimate the maximum radiation dose, for every type of cancer for which radiation doses are 

reconstructed, that could have been incurred under plausible circumstances by any member of the 

class; or (2) estimate the radiation doses of members of the class more precisely than a maximum dose 

estimate.

Although NIOSH found that it is not possible to completely reconstruct radiation doses for the 

proposed class, NIOSH intends to use any internal and external monitoring data that may become 

available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Therefore, dose reconstructions for individuals employed in

the Chemical Processing Plant at the Idaho National Laboratory during the period from January 1, 

1963 through December 31, 1974, but who do not qualify for inclusion in the SEC, may be performed 

using these data as appropriate.
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Health Endangerment Determination

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

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

of the proposed class from January 1, 1963 through December 31, 1974. 

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

establish that members of the proposed SEC class at the Idaho Chemical Processing Plant (CPP) 

between January 1, 1963 and December 31, 1974 were exposed to radiation during a discrete incident 

resulting in significant unmonitored exposures likely to have involved exceptionally high-level 

exposures. However, evidence indicates that some workers in the proposed class may have 

accumulated substantial chronic exposures through episodic intakes of radionuclides, combined with 

external exposures to gamma, beta, and neutron radiation. Consequently, NIOSH has determined that 

health was endangered for those workers covered by this evaluation who were employed for at least 

250 aggregated work days either solely under their employment or in combination with work days 

within the parameters established for other SEC classes.

For the workers at the INL site for the period of January 1, 1949 through December 31, 1970 

(excluding the CPP workers from January 1, 1963 through December 31, 1974, and those workers in 

the areas with reserved periods in this report), a health endangerment determination is not required 

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

the evaluated class.
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Table of Contents

1.0 Purpose and Scope ........................................................................................................................ 17

2.0 Introduction .................................................................................................................................. 17

3.0 SEC-00219 Idaho National Laboratory Class Definitions ........................................................... 18

3.1 Petitioner-Requested Class Definition and Basis ................................................................ 18

3.2 Class Evaluated by NIOSH ................................................................................................. 19

3.3 NIOSH-Proposed Class(es) to be Added to the SEC .......................................................... 19

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class ........................................................... 20

4.1 Site Profile Technical Basis Documents (TBDs) ................................................................ 23

4.2 ORAU Technical Information Bulletins (OTIBs) and Procedures ..................................... 23

4.3 Facility Employees and Experts .......................................................................................... 25

4.4 Previous Dose Reconstructions ........................................................................................... 26

4.5 NIOSH Site Research Database .......................................................................................... 27

4.6 Documentation and/or Affidavits Provided by Petitioners ................................................. 27

5.0 Radiological Operations Relevant to the Class Evaluated by NIOSH ......................................... 27

5.1 Idaho National Laboratory Plant and Process Descriptions ................................................ 27

5.1.1 Test Reactor Area (TRA) ........................................................................................ 31

5.1.1.1 Materials Test Reactor (MTR) ................................................................. 33

5.1.1.2 Engineering Test Reactor (ETR) .............................................................. 38

5.1.1.3 Advanced Test Reactor (ATR) ................................................................. 39

5.1.1.4 Reactivity Measurement Facility (RMF) .................................................. 42

5.1.1.5 Advanced Reactivity Measurement Facility No. 1 (ARMF-I) ................. 43

5.1.1.6 Advanced Reactivity Measurement Facility No. 2 (ARMF-II) ................ 43

5.1.1.7 Engineering Test Reactor Critical Facility (ERTC) ................................. 43

5.1.1.8 Advanced Test Reactor Critical Facility (ATRC) .................................... 44

5.1.1.9 Hot Cell Facility ....................................................................................... 45

5.1.1.10 TRA Gamma Facility ............................................................................... 48

5.1.1.11 Radiation Measurement Laboratory (RML) ............................................. 48

5.1.1.12 Radiochemistry Laboratories .................................................................... 48

5.1.1.13 HB-4 Crystal Spectrometer ...................................................................... 51

5.1.1.14 Fast (Neutron) Chopper ............................................................................ 51

5.1.2 Chemical Processing Plant (CPP) ........................................................................... 52

5.1.2.1 CPP-601, Main Processing Building ........................................................ 54

5.1.2.2 CPP-602 Laboratory ................................................................................. 62

5.1.2.3 CPP-603, Fuel Storage Building .............................................................. 63

5.1.2.4 CPP-604/605, Waste Disposal Building and Rare Gas Plant ................... 64

5.1.2.5 CPP-620, Chemical Engineering / High-Bay Laboratory ........................ 64

5.1.2.6 CPP-627, Remote Analytical Facility ...................................................... 64

5.1.2.7 CPP-630, Mass Spectrometry Facility ..................................................... 66

5.1.2.8 CPP-631, RaLa Off-Gas Filter Building .................................................. 66

5.1.2.9 CPP-633, Waste Calcining Facility .......................................................... 66
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5.1.2.10 CPP-637, Process Improvement Facility / Low-Bay Laboratory ............. 68

5.1.2.11 CPP-640, Hot Pilot Plant/Headend Process Plant .................................... 68

5.1.2.12 CPP-709, Eastside Service Waste (ESSW) Monitoring Station .............. 69

5.1.2.13 CPP-734, Westside Service Waste (WSSW) Monitoring Station ............ 69

5.1.2.14 High Level Liquid Waste Tank Farm ....................................................... 69

5.1.2.15 Calcined Solids Storage Facilities (CSSF) ............................................... 70

5.1.2.16 Auxiliary CPP Facilities (Non-Radiological) ........................................... 70

5.1.2.17 Overview: Fuel Reprocessing Operations ................................................ 70

5.1.2.18 Overview: Isotope Recovery .................................................................... 73

5.1.2.19 Overview: Waste Calcination ................................................................... 74

5.1.3 Test Area North (TAN) ........................................................................................... 76

5.1.3.1 Technical Support Facilities (TSF) .......................................................... 79

5.1.3.2 IET Facility ............................................................................................... 86

5.1.3.3 Flight Engine Test (FET)/Loss-Of-Fluid Test (LOFT) Facility ............... 90

5.1.3.4 Low Power Test Facility (LPTF) Area ..................................................... 92

5.1.4 Misc. Reactor Areas (ARA, OMRE, SPERT) ......................................................... 95

5.1.4.1 Auxiliary Reactor Area (ARA) ................................................................ 95

5.1.4.2 Organic Moderated Reactor Experiment (OMRE) ................................... 99

5.1.4.3 Special Power Excursion Reactor Tests (SPERT) ................................. 100

5.1.5 Central Facilities Area (CFA) ............................................................................... 103

5.1.6 Burial Ground ........................................................................................................ 107

5.2 Internal Radiological Exposure Sources from INL Operations ........................................ 113

5.2.1 Internal Radiological Exposure Sources from Test Reactor Area Operations ...... 113

5.2.1.1 Fission and Activation Products ............................................................. 114

5.2.1.2 Exotic Radionuclides .............................................................................. 114

5.2.1.3 Plutonium ............................................................................................... 114

5.2.1.4 Uranium .................................................................................................. 115

5.2.1.5 Thorium .................................................................................................. 115

5.2.2 Internal Radiological Exposure Sources from CPP Operations ............................ 115

5.2.2.1 Uranium .................................................................................................. 116

5.2.2.2 Neptunium .............................................................................................. 117

5.2.2.3 Plutonium ............................................................................................... 118

5.2.2.4 Fission and Activation Products ............................................................. 119

5.2.2.5 Barium/Lanthanum ................................................................................. 121

5.2.2.6 Thorium .................................................................................................. 121

5.2.2.7 Radioactive Noble Gases ........................................................................ 121

5.2.2.8 Iodine ...................................................................................................... 122

5.2.2.9 Tritium .................................................................................................... 123

5.2.3 Internal Radiological Exposure Sources from TAN Operations ........................... 123

5.2.3.1 Mixed Fission and Activation Products ................................................. 124

5.2.3.2 Radioiodines ........................................................................................... 124

5.2.3.3 Radioactive Noble Gases ........................................................................ 124

5.2.3.4 Plutonium ............................................................................................... 125

5.2.3.5 Neptunium .............................................................................................. 125

5.2.3.6 Uranium .................................................................................................. 125

5.2.3.7 Other Actinides ....................................................................................... 125

5.2.3.8 Other Radionuclides ............................................................................... 125
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5.2.4 Internal Radiological Exposure Sources from Misc. Reactor Areas Operations .. 126

5.2.4.1 Fission and Activation Products ............................................................. 126

5.2.4.2 Exotic Radionuclides .............................................................................. 127

5.2.4.3 Radioactive Noble Gases ........................................................................ 128

5.2.4.4 Iodine ...................................................................................................... 128

5.2.4.5 Uranium .................................................................................................. 129

5.2.5 Internal Radiological Exposure Sources from CFA Operations ........................... 129

5.2.5.1 Fission and Activation Products ............................................................. 130

5.2.5.2 Uranium .................................................................................................. 130

5.2.6 Internal Radiological Exposure Sources from Burial Ground Operations ............ 130

5.2.6.1 Fission and Activation Products ............................................................. 131

5.2.6.2 Plutonium ............................................................................................... 132

5.3 External Radiological Exposure Sources from INL Operations ....................................... 133

5.3.1 External Radiological Exposure Sources from Test Reactor Area Operations ..... 136

5.3.1.1 Photon ..................................................................................................... 136

5.3.1.2 Beta ......................................................................................................... 136

5.3.1.3 Neutron ................................................................................................... 136

5.3.2 External Radiological Exposure Sources from CPP Operations ........................... 136

5.3.2.1 Photon ..................................................................................................... 136

5.3.2.2 Beta ......................................................................................................... 139

5.3.2.3 Neutron ................................................................................................... 139

5.3.3 External Radiological Exposure Sources from TAN Operations .......................... 139

5.3.3.1 Photon ..................................................................................................... 139

5.3.3.2 Beta ......................................................................................................... 140

5.3.3.3 Neutron ................................................................................................... 140

5.3.4 External Radiological Exposure Sources from Misc. Reactor Areas Operations . 140

5.3.4.1 Photon ..................................................................................................... 140

5.3.4.2 Beta ......................................................................................................... 141

5.3.4.3 Neutron ................................................................................................... 141

5.3.5 External Radiological Exposure Sources from CFA Operations .......................... 141

5.3.5.1 Photon ..................................................................................................... 141

5.3.5.2 Beta ......................................................................................................... 142

5.3.6 External Radiological Exposure Sources from Burial Ground Operations ........... 142

5.3.6.1 Photon ..................................................................................................... 142

5.3.6.2 Beta ......................................................................................................... 142

5.4 Incidents ............................................................................................................................ 143

5.4.1 Incidents at the Test Reactor Area ......................................................................... 143

5.4.2 Incidents at the Chemical Processing Plant ........................................................... 148

5.4.3 Incidents at Test Area North .................................................................................. 151

5.4.4 Incidents at the Misc. Reactor Areas ..................................................................... 152

5.4.5 Incidents at Central Facilities Area ....................................................................... 153

5.4.6 Incidents at the Burial Ground .............................................................................. 153
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6.0 Summary of Available Monitoring Data for the Class Evaluated by NIOSH ........................... 154

6.1 Available INL Internal Monitoring Data ........................................................................... 154

6.2 Available INL External Monitoring Data .......................................................................... 158

6.3 Available Radiological Monitoring Data .......................................................................... 161

6.3.1 Available Radiological Monitoring Data for TRA ................................................ 161

6.3.2 Available Radiological Monitoring Data for CPP ................................................. 163

6.3.3 Available Radiological Monitoring Data for TAN ............................................... 171

6.3.4 Available Radiological Monitoring Data for Misc. Reactor Areas ....................... 174

6.3.5 Available Radiological Monitoring Data for CFA ................................................ 175

6.3.6 Available Radiological Monitoring Data for the Burial Ground ........................... 176

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by NIOSH .................................... 177

7.1 Pedigree of INL Data ......................................................................................................... 177

7.1.1 Internal Monitoring Data Pedigree Review ........................................................... 178

7.1.2 External Monitoring Data Pedigree Review .......................................................... 179

7.2 Evaluation of Bounding Internal Radiation Doses at INL ................................................ 179

7.2.1 Evaluation of Bounding Internal Doses for TRA .................................................. 181

7.2.1.1 Evaluation of Bounding Process-Related Internal Doses for TRA ........ 181

7.2.1.2 Methods for Bounding Operational Period Internal Dose for TRA ....... 185

7.2.2 Evaluation of Bounding Internal Doses for CPP ................................................... 186

7.2.2.1 Evaluation of Bounding Process-Related Internal Doses for CPP ......... 186

7.2.2.2 Methods for Bounding Operational Period Internal Dose for CPP ........ 191

7.2.3 Evaluation of Bounding Internal Doses for TAN .................................................. 195

7.2.3.1 Evaluation of Bounding Process-Related Internal Doses for TAN ........ 196

7.2.3.2 Methods for Bounding Operational Period Internal Dose for TAN ....... 197

7.2.4 Evaluation of Bounding Internal Doses for Misc. Reactor Areas ......................... 199

7.2.4.1 Evaluation of Bounding Process-Related Internal Doses for Misc. Reactor 

Areas ....................................................................................................... 199

7.2.4.2 Methods for Bounding Operational Period Internal Dose for Misc. 

Reactor Areas ......................................................................................... 201

7.2.5 Evaluation of Bounding Internal Doses for CFA .................................................. 202

7.2.5.1 Evaluation of Bounding Process-Related Internal Doses for CFA ........ 203

7.2.5.2 Methods for Bounding Operational Period Internal Dose for CFA ....... 204

7.2.6 Evaluation of Bounding Internal Doses for the Burial Ground ............................. 205

7.2.6.1 Evaluation of Bounding Process-Related Internal Doses for the Burial 

Ground .................................................................................................... 206

7.2.6.2 Methods for Bounding Operational Period Internal Dose for the Burial 

Ground .................................................................................................... 207

7.2.7 Internal Dose Reconstruction Feasibility Conclusion ........................................... 208

7.3 Evaluation of Bounding External Radiation Doses at INL ............................................... 210

7.3.1 Evaluation of Bounding Process-Related External Doses ..................................... 210

7.3.1.1 Electron and Photon Dosimetry Data ..................................................... 210

7.3.1.2 Neutron Dosimetry Data ......................................................................... 210

7.3.2 Evaluation of Bounding Ambient Environmental External Doses ........................ 211

7.3.3 Evaluation of Bounding Occupational Medical Doses ......................................... 211

7.3.4 Methods for Bounding External Dose at INL ....................................................... 212

7.3.4.1 Methods for Bounding Operational Period External Dose ..................... 212
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7.3.4.2 Methods for Bounding Ambient Environmental External Doses ........... 212

7.3.5 External Dose Reconstruction Feasibility Conclusion .......................................... 213

7.4 Evaluation of Petition Basis for SEC-00219 ..................................................................... 213

7.5 Other Potential SEC Issues Relevant to the Petition Identified During the Evaluation .... 214

7.6 Summary of Feasibility Findings for Petition SEC-00219 ................................................ 218

8.0 Evaluation of Health Endangerment for Petition SEC-00219 .................................................... 221

9.0 Class Conclusion for Petition SEC-00219 ................................................................................. 222

10.0 References .................................................................................................................................. 223

Attachment 1: Data Capture Synopsis ................................................................................................. 245

Tables

Table 4-1: INL Data Capture Visits ...................................................................................................... 21

Table 4-2: No. of INL Claims Submitted Under the Dose Reconstruction Rule .................................. 26

Table 5-1: INL Test Reactor Area (TRA) Reactors and Support Facilities .......................................... 32

Table 5-2: INL Chemical Processing Plant (CPP) Processing and Support Facilities .......................... 52

Table 5-3: CPP-601, Main Processing Building Cells .......................................................................... 61

Table 5-4: Notable Radiological Structures at the TAN ....................................................................... 78

Table 5-5: Nuclear Reactors that Operated at the LPTF ....................................................................... 94

Table 5-6: Dose Conversion Factors for Kr-85 and Xe-133 ............................................................... 116

Table 5-7: Fiscal Year-End Inventories of Radioactive Liquid Waste Activities ............................... 120

Table 5-8: Fiscal Year-End Inventories of Calcined Solid Waste Activities ...................................... 120

Table 5-9: Radioactive Noble Gases Released During Dissolution of Spent Fuel ............................. 121

Table 5-10: I-131 Release from C Cell Dissolving Operations During December 1956 .................... 122

Table 5-11: Radionuclides Measured in OMRE Coolant .................................................................... 127

Table 5-12: Example Air Effluent Inventory for SL-1 (ARA-II) ........................................................ 128

Table 5-13: Solid Radioactive Waste at INL Burial Ground .............................................................. 131

Table 5-14: Estimated Quantities of Radioactive Materials From Rocky Flats .................................. 132

Table 5-15: INL Beta-Gamma Dosimetry Systems History and LOD Summary ............................... 134

Table 5-16: INL Neutron Dosimetry Systems History and LOD Summary ....................................... 135

Table 5-17: Mixed Fission Product Gamma and Beta Emitters .......................................................... 135

Table 5-18: Mixed Fission Products from the RaLa Process .............................................................. 138

Table 6-1: In-Vitro Measurements for INL, 1953-1975 ...................................................................... 156

Table 6-2: In-Vivo Measurements for INL, 1961-1975 ....................................................................... 157

Table 6-3: Historic External Radiation Exposure Distribution for INLa ............................................. 159

Table 7-1: Feasibility Summary for TRA (1952-1970) ...................................................................... 181

Table 7-2: Special Samples Collected at the TRA .............................................................................. 183
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Table 7-3 [Radionuclide redacted] Bioassay Results for [Location and Role redacted] .................... 183

Table 7-4: MTR Smear Totals ............................................................................................................. 184

Table 7-5: Feasibility Summary for CPP (1953-1974) ....................................................................... 186

Table 7-6: CPP-Specific Internal Monitoring Records ....................................................................... 187

Table 7-7: CPP Permissible Airborne Contamination Levels ............................................................. 190

Table 7-8: CPP Contamination Monitoring ........................................................................................ 191

Table 7-9: Feasibility Summary for TAN (1955-1970) ...................................................................... 196

Table 7-10: Feasibility Summary for Misc. Reactor Areas (1955-1970) ............................................ 199

Table 7-11: Miscellaneous Reactor Area-Specific Internal Monitoring Results ................................ 200

Table 7-12: SPERT Whole Body Count Monitoring Frequency ........................................................ 201

Table 7-13: Feasibility Summary for Central Facilities Area (1949-1970) ........................................ 202

Table 7-14: Central Facilities Area-Specific Internal Monitoring Results ......................................... 203

Table 7-15: Feasibility Summary for the Burial Ground (1952-1970) ............................................... 205

Table 7-16: Example of Rocky Flats Waste Burial Schedule, February-August 1963 ....................... 207

Table 7-17: Summary of Feasibility Findings for SEC-00219 ........................................................... 218

Table 7-18: Summary of INL Feasibility Determinations by Operating Area (1949-1974) ............... 220

Figures

Figure 5-1: INL Location in Idaho ........................................................................................................ 28

Figure 5-2: Layout of INL Facilities ..................................................................................................... 29

Figure 5-3: Idaho National Laboratory Employment, 1951-1975 ......................................................... 30

Figure 5-4: Map Location, TRA ............................................................................................................ 31

Figure 5-5: Test Reactor Area (TRA) Complex, 1956 .......................................................................... 33

Figure 5-6: Inserting a Sample Rabbit in a Beam Port .......................................................................... 35

Figure 5-7: MTR Vertical Section Diagram (North-south Centerline Looking West) ......................... 36

Figure 5-8: A Portion of the MTR Canal .............................................................................................. 37

Figure 5-9: ETR Reactor ....................................................................................................................... 38

Figure 5-10: Advanced Test Reactor Complex (ATR in Center) .......................................................... 40

Figure 5-11: ATR Ground Floor ........................................................................................................... 41

Figure 5-12: Radiation Measurement Facility ....................................................................................... 42

Figure 5-13: The Engineering Test Reactor Critical Facility (ETRC) .................................................. 44

Figure 5-14: TRA Hot Cell #1 ............................................................................................................... 46

Figure 5-15: TRA Hot Cell #2 ............................................................................................................... 46

Figure 5-16: TRA Hot Cell #2 ............................................................................................................... 47

Figure 5-17: TRA Hot Cell Building, Internal Layout .......................................................................... 47

Figure 5-18: Lab 110, MTR Laboratory Wing ...................................................................................... 49

Figure 5-19: Alpha Lab 127, MTR Laboratory Wing ........................................................................... 49

Figure 5-20: Scientist Machining Thorium, Alpha Lab, 1964 .............................................................. 50

Figure 5-21: Preparation of Pure Pa-233, Alpha Cave .......................................................................... 51

Figure 5-22: Map of CPP Facilities ....................................................................................................... 53

Figure 5-23: Locations of CPP-601, CPP-602, CPP-630, and CPP-640 ............................................... 54

Figure 5-24: Perspective Drawing of CPP-601, Main Processing Building ......................................... 55

Figure 5-25: Operating Face of L-Cell .................................................................................................. 58

Figure 5-26: Shielded Glove Boxes in the Remote Analytical Facility ................................................ 65
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Figure 5-27: Map of Test Area North (TAN) ........................................................................................ 77

Figure 5-28: Map of the Technical Support Facilities (TSF) ................................................................ 80

Figure 5-29: Floor Plan of the A&M Building (TAN-607) .................................................................. 82

Figure 5-30: Floor Plan of Hot Cell Annex (TAN-633) ........................................................................ 85

Figure 5-31: Aerial Photograph of the IET Facility (facing east) ......................................................... 88

Figure 5-32: Map of the IET Facility .................................................................................................... 89

Figure 5-33: Map of the FET/LOFT Facility ........................................................................................ 91

Figure 5-34: Map of the LPTF Area ...................................................................................................... 92

Figure 5-35: Photograph of the LPTF Area (looking northwest) .......................................................... 93

Figure 5-36: ARA-I Hot Cell Wall with Viewing Windows and Remote-Handling Manipulators ...... 96

Figure 5-37: SL-1 in 1960 (Prior to 1961 Accident) ............................................................................. 97

Figure 5-38: ML-1 After Being Moved to ARA-III .............................................................................. 98

Figure 5-39: SPERT Area ................................................................................................................... 101

Figure 5-40: The SPERT-I Facility ..................................................................................................... 102

Figure 5-41: Map of the Central Facilities Area in 1961 with Buildings of Interest .......................... 104

Figure 5-42: ETR Reactor Vessel Arriving at CFA Gantry Crane in 1956 ........................................ 106

Figure 5-43: Road to the Burial Ground in 1961 ................................................................................ 107

Figure 5-44: Burial Ground, circa 1955 .............................................................................................. 108

Figure 5-45: Emplacement of Rocky Flats Drums in a Burial Ground Pit in 1958 ............................ 109

Figure 5-46: Mass Dumping of Rocky Flats TRU Drums in 1969 ..................................................... 110

Figure 5-47: January 1969 Flooding of Burial Ground ....................................................................... 111

Figure 5-48: Schematic of a Transuranic Storage Area ...................................................................... 112

Figure 5-49: Current Map of the Burial Ground and Associated Facilities ........................................ 112

Figure 5-50: Schematic of Moveable Glove Box (Initial Cause of Pu-240 Contamination) .............. 146

Figure 6-1: Radiation Survey of KAPL F/H-10 Cubicle in ETR, 4/24/64 .......................................... 162

Figure 6-2: Radiological Survey of MTR Alpha Labs June 6, 1962 ................................................... 163

Figure 6-3: Comparison of CPP-601 Access Corridor Contamination Surveys: 1961 ....................... 164

Figure 6-4: Comparison of CPP-601 Access Corridor Contamination Surveys: 1963 ....................... 165

Figure 6-5: Comparison of CPP-601 Access Corridor Contamination Surveys: 1965 ....................... 166

Figure 6-6: Comparison of CPP-601 Access Corridor Contamination Surveys: 1969 ....................... 167

Figure 6-7: Comparison of CPP-601 Access Corridor Contamination Surveys: 1970 ....................... 168

Figure 6-8: Example Contamination Survey for CPP-204 and CPP-207 in 1968 .............................. 170

Figure 6-9: Survey Schedule for Part of the TAN-607 Building ........................................................ 172

Figure 6-10: Example of Contamination Survey of TAN-633 (Hot Cell Annex) ............................... 173

Figure 6-11: Excerpt from SPERT Health Physics Monitoring Report, Aug. 1963 ........................... 174

Figure 6-12: Excerpt from CFA Health Physics Monthly Report, Jul. 1961 ...................................... 175

Figure 6-13: Safe Work Permit for RFP Waste Burial at INL Burial Ground, Apr. 1964 .................. 176

Figure 7-1: CPP Whole Body Count Frequency Schedule for 1963 ................................................... 188
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SEC Petition Evaluation Report for SEC-00219

1.0 Purpose and Scope

This report evaluates the feasibility of reconstructing doses for all employees who worked in any area 

of the Idaho National Laboratory from January 1, 1949 through December 31, 1970. It provides 

information and analyses germane to considering a petition for adding a class of employees to the 

congressionally-created SEC. 

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

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

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

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

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

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

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

2.0 Introduction

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

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

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

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

reconstructions.2 

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

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

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

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

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

estimate of the maximum radiation dose.

 

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

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

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

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

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

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

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

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

 

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

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

the detailed implementation guidelines available at http://www.cdc.gov/niosh/ocas.
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who were employed for at least 250 aggregated work days within the parameters established for the 

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

classes. 

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

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

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

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

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

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

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

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

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

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

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

3

 

3.0 SEC-00219 Idaho National Laboratory Class Definitions

The following subsections address the evolution of the class definition for SEC-00219, Idaho National 

Laboratory (INL). When a petition is submitted, the requested class definition is reviewed as 

submitted. Based on its review of the available site information and data, NIOSH will make a 

determination whether to qualify for full evaluation all, some, or no part of the petitioner-requested

class. If some portion of the petitioner-requested class is qualified, NIOSH will specify that class 

along with a justification for any modification of the petitioner’s class. After a full evaluation of the 

qualified class, NIOSH will determine whether to propose a class for addition to the SEC and will 

specify that proposed class definition. 

3.1 Petitioner-Requested Class Definition and Basis

Petition SEC-00219 was received on July 8, 2014, and qualified on September 16, 2014. The 

petitioner requested that NIOSH consider the following class: All employees who worked in any area 

of the Idaho National Laboratory from January 1, 1949 through December 31, 1970. 

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

accurate dose reconstruction over time is impossible for the INL workers in question. NIOSH deemed

the following handwritten statement sufficient to qualify SEC-00219 for evaluation: 

The petitioner indicated in Section F of the Special Exposure Cohort Petition – Form B, that there 

was no internal monitoring for plutonium, neptunium, or fission products (Petition, 2014). This 

serves as the basis for proposing that records and information at INL were inadequate for 

individual dose reconstruction. 

 

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

available at http://www.cdc.gov/niosh/ocas.
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The petitioner supplied a signed affidavit on August 29, 2014 (Affidavit, 2014). The affidavit 

stated “I affirm that to the best of my knowledge regarding the Idaho National Laboratory that 

there was no internal monitoring for Plutonium, Neptunium, or Fission Products during the 

employment years of 1947 to 1970. [Note: The year 1947 is in error; INL started in 1949 and the 

petitioner-requested class specified 1949.] 

Based on its INL research and data capture efforts, NIOSH determined that it has access to personnel 

bioassay monitoring (including urinalysis and fecal sampling), whole-body count data, and external 

dosimetry data for INL workers during the time period under evaluation. However, NIOSH also 

determined that internal monitoring records are not complete for all time periods or for all 

radionuclides. NIOSH concluded that there is sufficient documentation to support, for at least part of 

the requested time period, the petition basis that internal radiation exposures and radiation doses were 

not adequately monitored at INL, either through personal monitoring or area monitoring. The 

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

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

3.2 Class Evaluated by NIOSH

Based on its preliminary research, NIOSH accepted the petitioner-requested class. Therefore, NIOSH 

defined the following class for further evaluation: All employees who worked in any area of the Idaho 

National Laboratory from January 1, 1949 through December 31, 1970. 

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

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

which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSH-proposed class 

to be added to the SEC includes all employees of the Department of Energy, its predecessor agencies, 

and their contractors and subcontractors who worked at the Idaho National Laboratory in Scoville, 

Idaho, and were monitored for external radiation at the Idaho Chemical Processing Plant (CPP) (e.g., 

at least one film badge or TLD dosimeter from CPP) between January 1, 1963 and December 31, 1974 

for a number of work days aggregating at least 250 work days, occurring either solely under this 

employment, or in combination with work days within the parameters established for one or more 

other classes of employees in the Special Exposure Cohort. 

Based on its research, NIOSH concluded that the operating area and time period for which dose 

reconstruction is deemed infeasible is limited to the CPP. The SEC class start date was established as

January 1, 1963 because, in the face of increased alpha contamination, the CPP radiation protection 

program did not initiate a corresponding increase in the radiological monitoring of the workforce. The 

class end date was set to December 31, 1974 because there is documented evidence that the CPP 

radiation protection program did not start improving until 1975. Further NIOSH-ORAUT review and 

assessment will be necessary to determine if the program data from 1975 forward is adequate to 

support reconstructing dose with sufficient accuracy for this area.
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4.0 Data Sources Reviewed by NIOSH to Evaluate the Class

Idaho National Laboratory uses the Electronic Document Management System (EDMS) to manage 

records that are in an electronic format. EDMS is made of up of multiple client/server and web 

applications that run on several different operating systems and hardware to provide thousands of 

users with electronic documents. Preset as well as user-defined queries can be performed. The 

NIOSH and ORAUT Leads were provided EDMS accounts to allow for searches of non-classified 

documents during data capture trips in Idaho. EDMS was the primary resource in identifying and 

capturing documents of interest for the SEC-00219 Evaluation Report. 

As is standard practice, NIOSH also completed an extensive database and Internet search for 

information regarding Idaho National Laboratory. The database search included the DOE Legacy 

Management Considered Sites database, the DOE Office of Scientific and Technical Information 

(OSTI) SciTech Connect database, the OSTI Energy Citations database, and the Hanford Declassified 

Document Retrieval System. In addition to general Internet searches, the NIOSH Internet search 

included OSTI OpenNet Advanced searches, OSTI Information Bridge Fielded searches, Nuclear 

Regulatory Commission (NRC) Agency-wide Documents Access and Management (ADAMS) web 

searches, the DOE-National Nuclear Security Administration-Nevada Site Office-search, the Defense 

Technical Information Center database (DTIC), the DOE website, the DOE Comprehensive 

Epidemiologic Data Resource (CEDR), the DOE Environmental Management website, the DOE Oak 

Ridge Operation website, the Energy Employees Claimant Assistance Project (EECAP) website, the 

Hathitrust website, the Internet edition of the Health Physics Journal, the National Academies Press 

website, the Oak Ridge National Laboratory online library and report repository, the University of 

Hawaii, Manoa online collection, the University of North Texas online collection, the US Army Corps 

of Engineers website, and the US Transuranium and Uranium website. Attachment 1 contains a 

summary of INL documents. The summary specifically identifies data capture details and general 

descriptions of the documents retrieved. 

NIOSH and ORAUT staff went on ten data capture trips to assess and scan INL-related 

documentation. These ten trips resulted in the capture of a wide variety of relevant documentation, 

including 3,766 documents and 223,569 pages. Table 4-1 lists the INL data capture trips and the 

types of information obtained. 

In addition to the database and Internet searches listed above, NIOSH identified and reviewed 

numerous data sources to determine information relevant to determining the feasibility of dose 

reconstruction for the class of employees under evaluation. This included determining the availability 

of information on personal monitoring, area monitoring, industrial processes, and radiation source 

materials. The following subsections summarize the data sources identified and reviewed by NIOSH. 
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Table 4-1: INL Data Capture Visits

Dates Of Visit 

(Including Travel)

Volume Reviewed Documents 

Captured

Page 

Count

Brief Summary Of Data Capture Types

01/19/2015 -

01/22/2015

NA 143 789 Continuation of scanning of 10 boxes that were reviewed during site visit the week of 

January 5th.

01/05/2015 -

01/09/2015

93 Boxes and 1 DVD

552 20,601

Engineering Test Reactor (ETR) film badges including neutron, ETR primary coolant, Idaho 

Chemical Processing Plant (ICCP), Test Area North (TAN) decontamination logs, disposal 

of leaking radium sources, selected activity reports of the health physics and monitoring 

branches, Chemical Processing Plant (CPP) air samples and floor plan surveys, thyroid and 

urinalysis data for Ferguson Construction, CPP incident reports, excerpts from CPP health 

physics logs, 1956 radiation shipment survey samples, Central Facilities Area (CFA) daily 

health physics log sheets, safe work permits for hot cell entries, ETR contamination events, 

incident study of CPP, ETR safe work permits, Materials Testing Reactor (MTR) shipment 

records, MTR routine samples to CPP, MTR contamination survey data, CPP safe work 

permits, TAN radioactive shipments, waste management, annual report on reactor fuels, 

smear surveys, general decontamination procedures, MTR radioactive discharge records, 

waste disposal request, ICPP health physics monthly reports, and health and safety monthly 

report.

12/8/2014 -

12/12/2014

79 Boxes, 6 DVDs

426 33,599

Radioactive material shipment records, Central Facilities Area (CFA) health physics logs, 

urine sample results, Special Power Excursion Reactor Test (SPERT) health physics logs, 

Idaho Chemical Processing Plant (ICCP) safe work permits, radioactive and non-radioactive 

laundry shipments, Center for Advanced Modeling and Simulation (CAMS) filter counts, 

incident reports for reactor, thyroid counts, film reports, health physics survey sheets, fallout 

log sheets, daily station reports, SPERT film badge reports, safe work permits, Materials 

Testing Reactor (MTR) safe work permits, MTR radiation survey data, personnel 

monitoring practices, MTR gamma facility, SPERT health physics monthly reports, eye 

exams for neutron damage, Engineering Test Reactor (ETR) water samples, ETR smear 

surveys, Chemical Processing Plant (CPP) air samples, CPP and MTR health physics 

logbooks, burial ground photographs, radioactive shipment incidents, contamination 

incidents, and thyroid counts for 1958.

11/17/2014 -

11/21/2014

49 Boxes

282 9,677 Trip was in support of SEC-00219. Documents included reactor smear surveys, shipment 

records, safe work permits, monthly reports, waste reports, and health physics logbooks. 

10/27/2014 -

10/31/2014

107 Boxes

265 34,156

Trip was in support of SEC-00219. Documents included safe work permits, health physics 

logbooks, and personnel exposure records. Documents are currently going classification 

review and will be released after they are cleared.

09/29/2014 -

10/01/2014

NA 106 7,792 Completed scanning documents from 15 boxes from the previous data capture to the site.
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Table 4-1: INL Data Capture Visits

Dates Of Visit 

(Including Travel)

Volume Reviewed Documents 

Captured

Page 

Count

Brief Summary Of Data Capture Types

09/08/2014 -

09/11/2014

67 Boxes

314 2,804

Radiological incident reports, contamination control letters, criticality safety procedure, 

medical and urinalysis records, selected reports from possible internal intakes, neutron 

historical records, whole body count employee records and contamination control upgrade 

plants.

06/04/2012 - 06 

/14/2012

165 Boxes of Records 

and one box of 

microfiche

1,176 81,027

Air sample data, incident reports, personnel dosimetry files, health physics procedures, 

radiation exposure reports, dosimetry data, monthly health physics reports, and air 

monitoring data.

05/07/2012 -

05/11/2012

71 Boxes

302 22,870

Neutron documents, visitor dosimetry reports, airborne release summaries, air sample data, 

environmental air monitoring data, personnel contamination, quarterly environmental 

reports, environmental monitoring program reports, positive whole body count data, 

contamination control, particle information, health physics logbook, dosimetry procedures 

for film dosimetry.

04/02/2012 -

04/06/2012

29 Boxes

200 26,489 A site tour was conducted. Also, 29 boxes were reviewed. The selected documents were 

scanned and are awaiting classification review.

TOTALS: 

660 boxes, 7 DVDs, 

and 1 box of 

microfiche

3,766 239,804 N/A
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4.1 Site Profile Technical Basis Documents (TBDs)

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

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

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

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

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

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

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

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

operations or related topics/operations at other sites: 

• Idaho National Laboratory and Argonne National Laboratory-West – Introduction, ORAUTTKBS-0007-1, Rev. 03; March 12, 2010; SRDB Ref ID: 79862 

• Idaho National Laboratory and Argonne National Laboratory-West – Site Description, ORAUTTKBS-0007-2, Rev. 04; August 2, 2010; SRDB Ref ID: 84106 

• Idaho National Laboratory and Argonne National Laboratory-West – Occupational Medical 

Dose, ORAUT-TKBS-0007-3, Rev. 02; December 21, 2009; SRDB Ref ID: 77829 

• Idaho National Laboratory and Argonne National Laboratory-West – Occupational 

Environmental Dose, ORAUT-TKBS-0007-4, Rev. 02; January 8, 2010; SRDB Ref ID: 78635 

• Idaho National Laboratory and Argonne National Laboratory-West – Occupational Internal 

Dose, ORAUT-TKBS-0007-5, Rev. 03; March 2, 2010; SRDB Ref ID: 79571 

• Idaho National Laboratory and Argonne National Laboratory-West – Occupational External 

Dose, ORAUT-TKBS-0007-6, Rev. 03; April 19, 2011; SRDB Ref ID: 94104 

4.2 ORAU Technical Information Bulletins (OTIBs) and Procedures

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

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

Procedure provides specific requirements and guidance regarding EEOICPA project-level activities, 

including preparation of dose reconstructions at particular sites or categories of sites. NIOSH 

reviewed the following OTIBs and procedures as part of its evaluation: 

• OTIB: Internal Dosimetry Organ, External Dosimetry Organ, and IREP Model Selection by ICD9 Code, ORAUT-OTIB-0005 Rev 05; December 20, 2012; SRDB Ref ID: 121336 

• OTIB: Dose Reconstruction from Occupational Medical X-Ray Procedures, ORAUT-OTIB-0006, 

Rev. 04; June 20, 2011; SRDB Ref ID: 98147

• OTIB: Technical Information Bulletin: Tritium Calculated and Missed Dose Estimates, ORAUTOTIB-0011 Rev 00; June 29, 2004; SRDB Ref ID: 19430 
SEC-00219 03-12-15 Idaho National Laboratory

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

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

19432 

• OTIB: Bulletin: Interpretation of Dosimetry Data for Assignment of Shallow Dose, ORAUTOTIB-0017 Rev 01; October 11, 2005; SRDB Ref ID: 19434 

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

• OTIB: Technical Information Bulletin: Assignment of Missed Neutron Doses Based on Dosimeter 

Records, ORAUT-OTIB-0023 Rev 01; May 14, 2008; SRDB Ref ID: 43937 

• OTIB: Application of Internal Doses Based on Claimant-Favorable Assumptions for Processing 

as Best Estimates, ORAUT-OTIB-0033 Rev 00; April 20, 2005; SRDB Ref ID: 19457 

• OTIB: Estimating Doses for Plutonium Strongly Retained in the Lung, ORAUT-OTIB-0049 Rev 

01 PC-2; November 29, 2010; SRDB Ref ID: 90666 

• OTIB: Technical Information Bulletin: Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gross Gamma Analyses, ORAUT-OTIB-0054 Rev 02; March 6, 

2014; SRDB Ref ID: 130852 

• OTIB: Technical Information Bulletin: Internal Dose Reconstruction, ORAUT-OTIB-0060 Rev 

00; February 6, 2007; SRDB Ref ID: 29984 

• OTIB: Guidance on Assigning Occupational X-Ray Dose under EEOICPA for X-Rays 

Administered Off Site, ORAUT-OTIB-0079 Rev 00; January 3, 2011; SRDB Ref ID: 89563 

• OTIB: Technical Information Bulletin: Dose Reconstruction Method for Chronic Lymphocytic 

Leukemia, ORAUT-OTIB-0082 Rev 00 PC-1; December 20, 2012; SRDB Ref ID: 121337 

• Procedure: Occupational On-Site Ambient Dose Reconstruction for DOE Sites, ORAUT-PROC0060 Rev 01; June 28, 2006; SRDB Ref ID: 29986 

• Procedure: Occupational Medical X-Ray Dose Reconstruction for DOE Sites, Rev 03, ORAUTPROC-0061 Rev 03; March 3, 2010; SRDB Ref ID: 79758 

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4.3 Facility Employees and Experts

To obtain additional information, NIOSH interviewed nine former INL employees:

NOTE: The interviews below are not finalized until they have been reviewed by the interviewees. 

That process is on-going at the time of the issuance of this evaluation report. 

• Personal Communication, 2014a, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by NIOSH and SC&A; November 2014; SRDB Ref ID: 141472

• Personal Communication, 2014b, Personal Communication with former INL worker; In-person 

interview, 10:00 A.M. at Idaho Falls, Idaho by ORAU Team, ABRWH, SC&A, and NIOSH; 

November 17, 2014; SRDB Ref ID: 141476

• Personal Communication, 2014c, Personal Communication with former INL worker; In-person 

interview, 13:45 P.M. at Idaho Falls, Idaho by ORAU Team, ABRWH, SC&A, and NIOSH; 

November 17, 2014; SRDB Ref ID: 141473

• Personal Communication, 2014d, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, Board, SC&A, and NIOSH; November 18, 2014; 

SRDB Ref ID: 141475

• Personal Communication, 2014e, Personal Communication with former INL worker; In-person 

interview, 13:15 P.M. at Idaho Falls, Idaho by ORAU Team, Board, SC&A, and NIOSH; 

November 18, 2014; SRDB Ref ID: 141477

• Personal Communication, 2014f, Personal Communication with former INL worker; In-person 

interview, 11:00 A.M. at Idaho Falls, Idaho by SC&A and NIOSH; November 18, 2014; SRDB 

Ref ID: 141478

• Personal Communication, 2014g, Personal Communication with former INL worker; In-person 

interview, 17:45 P.M. at Idaho Falls, Idaho by ORAU Team, Board, SC&A, and NIOSH; 

November 18, 2014; SRDB Ref ID: 141480

• Personal Communication, 2014h, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, SC&A, and NIOSH; November 20, 2014; SRDB 

Ref ID: 141479

• Personal Communication, 2014i, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team and NIOSH; December 11, 2014; SRDB Ref ID: 

141471
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4.4 Previous Dose Reconstructions

NIOSH reviewed its NIOSH DCAS Claims Tracking System (referred to as NOCTS) to locate

EEOICPA-related dose reconstructions that might provide information relevant to the petition 

evaluation. Table 4-2 summarizes the results of this review. (NOCTS data available as of February 2, 

2015) 

Table 4-2: No. of INL Claims Submitted Under the Dose Reconstruction Rule

Description Totals

Total number of claims submitted for dose reconstruction 1854

Total number of claims submitted for energy employees who worked during the period under 

evaluation (January 1, 1949 through December 31, 1974). Use the NIOSH-evaluated class 

timeframe here and below. 1074

Number of dose reconstructions completed for energy employees who worked during the period

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

Department of Labor for final approval). 956

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

evaluated class definition 731

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

evaluated class definition 711

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

records could be obtained for the employee. The Computer Assisted Telephone Interviews (CATI) 

provided some information that is useful for dose reconstruction, including work locations, hours 

worked, incidents (e.g., fires, radioactive releases, and spills), and hazards encountered. Of the 1,074 

claims submitted with employment during the period under evaluation, 731 (68%) have internal 

monitoring data available and 711 (66%) have external monitoring data available. 
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4.5 NIOSH Site Research Database

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

assessment of the evaluated class. There are currently 9,487 documents in this database identified as 

pertaining to Idaho National Laboratory. These documents were evaluated for their relevance to this 

petition. The documents include historical background on data types, monitoring, and program 

descriptions found (e.g., dust sampling, air monitoring, urinalysis data, radiological control program, 

medical monitoring, process materials, and process description).

4.6 Documentation and/or Affidavits Provided by Petitioners

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

petitioner:

The petitioner supplied a signed affidavit on August 29, 2014 (Affidavit, 2014). The affidavit 

stated: I affirm that to the best of my knowledge regarding the Idaho National Laboratory that 

there was no internal monitoring for Plutonium, Neptunium, or Fission Products during the 

employment years of 1947 to 1970. [Note: The year 1947 is in error; the INL site began in 1949 

and the petitioner-requested class specified 1949 as the starting year.]

5.0 Radiological Operations Relevant to the Class Evaluated by 

NIOSH

The following subsections summarize both radiological operations at Idaho National Laboratory from 

January 1949 to December 1970 and the information available to NIOSH to characterize particular 

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

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

and information describing processes through which radiation exposures may have occurred and the 

physical environment in which they may have occurred. The information included within this 

evaluation report is intended only to be a summary of the available information. 

5.1 Idaho National Laboratory Plant and Process Descriptions 

The Idaho National Laboratory was known by several names throughout its history [i.e., National 

Reactor Testing Station (1949–1973), Idaho National Engineering Laboratory (1974–1996), Idaho 

National Engineering and Environmental Laboratory (1997–2004), and Idaho National Laboratory 

(2005–Present)]. For the purposes of this evaluation report, this site will be referred to as the Idaho 

National Laboratory or INL (with the exception of source document titles) throughout the rest of this 

evaluation report. 

Idaho National Laboratory is an 890-square-mile complex located in the high desert of eastern Idaho, 

west-northwest of the city of Idaho Falls (Figure 5-1). In 1949, the Atomic Energy Commission 

(AEC) established the National Reactor Testing Station on the site of a 1940s United States Navy 

bombing and artillery range. INL was the primary nuclear reactor development laboratory in the 

United States. Over 100 reactor concepts were conceived and tested there. Between 1953 and 1992, 
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the Idaho Chemical Processing Plant (CPP) at INL reprocessed spent nuclear fuel from naval 

propulsion, test, and research reactors to recover enriched uranium for reuse in nuclear fuel and 

nuclear weapons production. Other facilities at INL also conducted various nuclear weapons research 

and development activities. 

Figure 5-1: INL Location in Idaho

There are six INL operating areas that will be addressed in this evaluation report. Figure 5-2 shows 

the layout of the INL facilities. 

1. Test Reactor Area (TRA) – Section 5.1.1 

2. Chemical Processing Plant (CPP) – Section 5.1.2 

3. Test Area North (TAN) – Section 5.1.3 

4. Misc. Reactor Areas (ARA, OMRE, SPERT) – Section 5.1.4 

5. Central Facilities Area (CFA) – Section 5.1.5 

6. Burial Ground – Section 5.1.6 
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Figure 5-2: Layout of INL Facilities

In Figure 5-2, the blue areas are considered INL, the green areas are considered Argonne National 

Laboratory-West (ANL-W), and the red is the Naval Reactors Facility (NRF), which is not a covered 

facility under EEOICPA.
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Figure 5-3 shows the employment levels at INL from 1951 through 1975. After reviewing the 

available INL documentation, NIOSH has concluded that definitive historical site population numbers 

are elusive. There is no consistent source of employment information and the various sources count 

employees differently. INL’s own historical overview of site exposure history points out that there is 

little evidence of real effort to identify and preserve historic data. Those INL authors experienced 

considerable difficulty obtaining total counts before 1974, much less a breakdown by individual 

contractor (Exposure History, 1993). Most sources count only the AEC and contractor personnel 

administered by IDO; others may count non-IDO contract employees (e.g., ANL-W), and sometimes 

visitors are counted. These counts do not appear to include workers employed by minor 

subcontractors (e.g., construction workers), which only appear to be included when visitors are 

counted. Counting visitors can greatly expand the population count (e.g., for 1961, from 4,920 to 

25,580). Thus, INL employment figures are best viewed as “ballpark” numbers that reflect AEC and 

primary contractor personnel.

Source: Exposure History, 1993

Figure 5-3: Idaho National Laboratory Employment, 1951-1975
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5.1.1 Test Reactor Area (TRA)

ATTRIBUTION: Section 5.1.1 was completed by Mike Mahathy, Oak Ridge Associated Universities. 

All conclusions drawn from the data regarding the Test Reactor Area were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the Test Reactor Area are explained in the associated text.

The Test Reactor Area (TRA) is approximately five miles north of the Central Facilities Area. Figure 

5-4 shows the location of TRA relative to several other INL operating areas. 

Figure 5-4: Map Location, TRA

Eight reactors were built and operated in the TRA (Stacy, 2000): 

• High-flux Reactors Used for Materials Testing

- Materials Test Reactor (MTR)

- Engineering Test Reactor (ETR)

- Advanced Test Reactor (ATR)

• Low-power Reactors Used for Reactivity Measurements

- Reactivity Measurement Facility (RMF)

- Advanced Reactivity Measurement Facility No. 1 (ARMF-1) 

- Advanced Reactivity Measurement Facility No. 2 (ARMF-2) 

- Engineering Test Reactor Critical (ETRC)

- Advanced Test Reactor Critical (ATRC)
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Over the course of MTR and ETR operations, more than 250 radionuclides were either irradiated or 

produced from irradiation. The following support and laboratory facilities were also located in the 

Test Reactor Area: 

• TRA Hot Cells

• Gamma Facility

• Radiation Measurement Laboratory

• Radiochemistry Laboratories 

• HB-4 Crystal Spectrometer

• Fast (Neutron) Chopper 

All personnel who entered the TRA were required to wear dosimetry badges, and those who worked 

in or near radiological control areas were issued pocket ionization chambers as well. Personnel who 

worked in radiological control areas were on a routine bioassay program and received routine 

whole-body counts. Table 5-1 shows the TRA facilities and their years of first and last use. These 

facilities are discussed in the following subsections.

Table 5-1: INL Test Reactor Area (TRA) Reactors and Support Facilities

Facility Name Building No. First Used Last Used Facility Type

Materials Test Reactor (MTR) TRA-603 1952 1970 Reactor

Engineering Test Reactor (ETR) TRA-642 1957 1981 Reactor

Advanced Test Reactor (ATR) TRA-670 1967 In use Reactor

Reactivity Measurement Facility (RMF) TRA-603 1954 1962 Reactor

Advanced Reactivity Measurement Facility No. I (ARMF-I) TRA-660 1960 1974 Reactor

Advanced Reactivity Measurement Facility No. 2 (ARMF-II); 

renamed Coupled Fast Reactivity Measurement Facility 

(CFRMF)

TRA-660 1962 1991 Reactor

Engineering Test Reactor Critical Facility (ETRCF) TRA-654 1957 1982 Reactor

Advanced Test Reactor Critical Facility (ATRC) TRA-670 1964 In use Reactor

Hot Cell Facility TRA-632 1954 In use Research

Gamma Facility TRA-641 1955 ND Research

Radiation Measurement Laboratory (RML) TRA-604 1952 In use Sampling/

Research 

Radiochemistry Laboratories TRA-604 1950 In use Sampling/

Research

Alpha Laboratories TRA-652 1961 1970 Sampling/

Research

HB-4 Crystal Spectrometer TRA-603 1952 1970 Research

Fast (Neutron) Chopper TRA-665 1952 1970 Research

ND = Date of last use not determined
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Figure 5-5 shows an aerial view of the Test Reactor Area taken in 1956. The large building in the left 

foreground is the MTR building; the large building to the right in the background is the ETR Building. 

The attached wing at the bottom contains the Alpha Laboratories. 

Figure 5-5: Test Reactor Area (TRA) Complex, 1956

5.1.1.1 Materials Test Reactor (MTR)

The MTR was the original reactor at the TRA and the second reactor to be operated at INL. This 

water-cooled, water-moderated reactor used enriched uranium fuel and used beryllium and graphite as 

a reflector. Fuel assemblies were plate-type and fabricated from aluminum (Stroschein, 1967). The 

MTR operated at a power level of 30 megawatts until September 1955, when output was increased to 

40 megawatts. It supplied a high-neutron flux in support of a reactor development program that 

irradiated potential reactor fuels and structural materials (Stacy, 2000). Shielding was used to prevent 

personnel exposures from neutron and gamma radiation. 

The MTR was designed to test concepts and designs for future reactors. The high density of 

epithermal neutrons in the MTR permitted the extension of the practical range of the crystal 

spectrometer to higher neutron energy ranges. The properties of fissionable isotopes in the 

lower-resonance energy regions could be investigated and their nuclear constants more accurately 

determined. The MTR provided the opportunity for reactions of higher order with the capture of two 

or more neutrons in succession by the same nucleus in significant quantities of rare isotopes, thereby 

permitting the study of their nuclear constants. Production of radioisotopes in amounts and of specific 

activities previously not readily available was done with the MTR. Tests of potential nuclear fuels 
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were conducted under simulated operating conditions of temperature, pressure, and environmental 

factors (Stroschein, 1967). 

The design of the MTR provided researchers with several options for sample irradiation. Pneumatic 

ports on the reactor top enabled the insertion of capsules for irradiation into a graphite region around 

the core. A hydraulic rabbit system underneath the reactor enabled the insertion of specimens and 

their discharges to the canal during reactor operation. Horizontal and angular beam holes made it 

possible to perform cross-section measurements and other physics research experiments (Figure 5-6). 

A component, referred to as a discharge chute, was installed in the reactor vessel and used to 

discharge spent fuel assemblies, control rods, irradiated slugs, and components to the canal 

(Stroschein, 1967). The high-flux radiation fields available in the MTR made it possible to accelerate 

the screening of test materials. The MTR contributed to the design of various types of reactors, 

including pressurized-water, organic-moderated, liquid-metal-cooled, and others (Stacy, 2000, PDF p. 

279).

In August 1958, the MTR became the first reactor to operate using Pu-239 as fuel with about 4 kg 

Pu-240 as a convertible poison (Keller, 1958a). The MTR was again fueled with Pu-239 with 5.6 kg 

Pu-240 in 1970 for limited experiments as part of the Hanford Phoenix Fuel Program (Hofmann, 

1965; MTR, 1967; Davis, 1970). The plutonium cores demonstrated that a plutonium-fueled, watermoderated reactor could be controlled satisfactorily. The MTR logged more than 125,000 operating 

hours in more than 19,000 neutron irradiations. In August 1970, the MTR was again brought to 

power for a twenty-four hour run to irradiate one thousand biological samples for iodine analysis

which was the last operation (Stacy, 2000).
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Figure 5-6: Inserting a Sample Rabbit in a Beam Port

The MTR reactor building was built to enclose the reactor structure and canal. It also furnished space 

for experimental facilities on the main floor level and in the basement. The control room, operations 

personnel offices, and electrical equipment were housed on two balconies above the main floor. The 

building was designed to operate at a positive pressure of 1-1/2 in. of water to reduce entrance of dust 

into the building; but, due to building additions and increased use of ventilating air, atmospheric 

pressure prevailed by 1967 (Stroschein, 1967). Building air was drawn through filters into the reactor 

structure and was used to cool the graphite reflector. Building walls were comprised of pre-cast 

insulated concrete slabs bolted to concrete-encased structural columns. They were designed to be 

blown outward in the event of a high pressure surge in the building. The flat roof was constructed of 

pre-cast concrete panels laid on steel purlins covered with foam-glass insulating blocks and other 

roofing material. The main floor of the reactor building was reinforced concrete. The floor area 

immediately surrounding the reactor structure was three feet thick; the floor surrounding this area and 

extending to the building wall was one foot thick. 
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Figure 5-7 shows a vertical section diagram of the MTR reactor through the north-south centerline 

looking west. 

Figure 5-7: MTR Vertical Section Diagram (North-south Centerline Looking West)
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A canal was designed and constructed as a shielding medium. The use of water in a canal 

arrangement allowed for direct visual contact of radioactive components, and facilitated the use of 

simple tools in the handling of those components. The canal contained equipment for handling the 

assemblies discharged from the reactor. The canal provided space for storage of spent fuel 

assemblies, irradiated materials, and internal reactor parts. The canal walls were originally covered 

with 8 in. by 16 in. white glazed structural tile four-inches thick. The bottom was lined with four 

inches of white concrete. A stainless steel liner was added after excessive leakage was experienced. 

A flow of 20 gallons per minute of de-mineralized water was maintained through the canal to avoid an 

excessive build-up of radioactivity and turbidity. With the original construction, water from the canal 

overflowed a weir in the east end of the canal and flowed through a discharge pipe to the canal sump. 

However, in time, an overflow pipe was installed west of the canal isolation gate. The use of the 

overflow weir was discontinued (Stroschein, 1967). Figure 5-8 shows the canal.

Figure 5-8: A Portion of the MTR Canal

There was considerable maintenance work performed on the canal during shutdowns. This work 

required constant monitoring by Health Physics. The spread of contamination was a constant hazard 

during canal work; Health Physics staff was trained to recognize conditions that could have possibly 

caused contamination and the steps to prevent it (Stroschein, 1967). Health Physics logbooks that are 

available to NIOSH demonstrate the oversight of canal operations. 
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5.1.1.2 Engineering Test Reactor (ETR)

The ETR was designed and constructed for performing engineering tests of fuel elements and nuclear 

plant components. It was placed into operation in 1957. At that time, the ETR was the largest and 

most advanced nuclear materials test reactor in the world. The ETR provided larger test spaces than 

the MTR and a more intense neutron flux. ETR fuel, coolant, and moderator materials were evaluated 

under environments similar to those of power reactors. Several experimental loop facilities were 

designed to test fuels for the Aircraft Nuclear Propulsion (ANP) project and the U.S. Navy’s fuel 

development program. 

The ETR reactor building was 136 feet long and 112 feet wide. It extended 58 feet above grade and 

38 feet below grade to the basement floor. The high bay area above the first floor had a clear height 

of 47 feet to the bottom of the roof truss structure. The first floor was at the same floor elevation as 

the MTR main floor. A photo of the reactor housing shown from the first floor showing clean 

conditions is given in Figure 5-9. 

The ETR canal extended westward from the reactor structure. The process-water-pipe tunnel 

extended eastward from the reactor structure under the first floor. The console floor was about 22 feet

below the main floor and served a dual purpose as a shielding roof for experimental cubicles at 

basement level, and as a working level for experimental consoles and equipment. The walls of the 

pipe tunnel, the biological shield, and the canal walls extending from the first floor to the console 

floor divide the console area into halves connected only at the west end by a 9-foot 6-inch wide 

corridor. The basement floor was approximately 38 feet below the main floor resting on compacted

fill (Stroschein, 1967).

Figure 5-9: ETR Reactor
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A working canal provided ample working room for removal of experiments from the reactor tank. 

Fuel elements and experimental sections were transferred from the reactor into the working canal 

through the discharge chute in the reactor vessel transition piece. That was accomplished by lowering 

the fuel element into a transfer tube located in the canal under the discharge chute. The transfer tube 

(which was hydraulically operated) was pivoted to permit an operator to secure the element from the 

canal by a hook tool. The operator then led the element to the storage racks. Fuel elements had a 

tendency to float when transferred through the water with any speed, which presented an additional 

radiation hazard. The canal working platform was permanently attached to the east end of the canal 

parapet allowing access to the working area of the canal immediately over the reactor transfer tube. 

The platform was constructed in two sections on either side of the canal with an open working area in 

between the sections. Each section was served with a stairway from the first floor. An underwater 

saw was located on the south wall of the working canal to cut off fuel element end-boxes and to cut 

in-pile piping into disposable lengths. A capsule reloading tray, flux wire storage grid, and an 

underwater periscope used for capsule inspection were located by the south canal wall (Stroschein, 

1967).

The storage canal had storage space for irradiated fuel slugs, experimental equipment, fuel elements, 

and reactor core components. This storage was laid out by alphabetical grids. The north half of the 

storage canal was designated for storage of non-fissionable core components while fuel assemblies 

were stored in cadmium-lined storage grids. Spent fuel elements were stored in 36-hole, cadmiumlined storage grids to cool before being cut and shipped to either the Gamma Facility or the Chemical 

Processing Plant. The south half of the canal storage area was used for storing long in-pile loop 

experimental samples and other bulky items (Stroschein, 1967).

There was potential for external exposure to workers in the reactor area during shutdown, during 

changes of loop, and as experiment samples were taken. 

Potential for internal exposure from airborne fission products was minimal during normal reactor 

operation. There was some potential for internal exposures during shutdown because airborne fission 

products were often released when the reactor top was removed. There was also some potential for 

internal exposure from activation products during maintenance work in the reactor subpile room and 

in the experiment cubicles.

5.1.1.3 Advanced Test Reactor (ATR)

The ATR is the newest materials testing reactor to be built in the Test Reactor Area, going into service 

in July 1967; it is still operational. Its design is based on ideas developed through use of the MTR and 

ETR. The ATR simulates the environment in a power reactor to study the effect of radiation on steel, 

zirconium, and other materials. The reactor produces an extremely high neutron flux that makes it 

ideal for materials testing. Target materials are exposed to the neutron flux to test their durability in 

an environment of high temperature, high pressure, and high gamma fields. Data that normally would 

require years to gather from ordinary reactors can be obtained in weeks or months from the ATR. 
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The ATR can operate up to a power level of 250 MW. Its unique four-lobed design can deliver a wide 

range of power levels to nine main test spaces or loops. Each loop has its own distinct environment 

apart from that of the main reactor core. Smaller test spaces around the loops enable additional tests. 

The ATR also produces radioisotopes for use in medicine, industry, and other research (Stacy, 2000, 

PDF p. 274).

The ATR is located in the northwest portion of the Test Reactor Area complex. Figure 5-10 shows a 

modern-day aerial photo of the ATR complex. The reactor building covers an area of approximately 

200 feet by 200 feet, while the substructure extends approximately 60 feet below grade.

Figure 5-10: Advanced Test Reactor Complex (ATR in Center)
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On the ground floor, the ATR reactor room is 88 feet by 100 feet and occupies the south central 

portion of the building (Figure 5-11). 

Figure 5-11: ATR Ground Floor
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5.1.1.4 Reactivity Measurement Facility (RMF)

The Reactivity Measurement Facility (RMF) was a very-low-power reactor in the east end of the 

MTR canal that operated at a power level of 100 or 200 W. It was installed in August 1954 and 

placed in use by the end of 1954 (Notes, 2015). Water was used as the moderator, reflector, and 

shield. The RMF was designed to measure reactivity changes in materials irradiated in the MTR or 

ETR. The RMF was used to assay new and spent fuel elements and to assist in experiment scheduling 

by evaluating reactivity losses and flux depression caused by in-pile apparatus (Stacy, 2000, PDF p. 

280). Figure 5-12 shows the RMF over the top of the MTR canal. It was replaced with the Advanced 

Reactivity Measurement Facility No. 1.

Figure 5-12: Radiation Measurement Facility

Specific RMF applications included precision measurements of reactivity that were employed when 

investigating: gross fission-product cross-section as a function of operating time in a high-flux reactor; 

fuel burn-up and burnable poison studies; and determination of pile cross-sections and resonance 

integrals of special materials (e.g., zirconium alloys, dysprosium, and hafnium). The RMF measured 

the rate of U-233 build-up in thorium after MTR irradiation. The high cross-section fission product 

Xe-135 was readily observed as it was formed by decay of I-135, and subsequently, as it disappeared 

by decay to Cs-135 (Phillips, 1962).
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5.1.1.5 Advanced Reactivity Measurement Facility No. 1 (ARMF-I)

ARMF-I was a small pool reactor in Building TRA-660 located east of the MTR. It was used to 

determine the nuclear characteristics of reactor fuels and other materials for testing in the MTR. 

ARMF-I operated from October 1960 through 1974 (Stacy, 2000, PDF p. 274).

5.1.1.6 Advanced Reactivity Measurement Facility No. 2 (ARMF-II)

ARMF-II was a pool reactor in the opposite end of the tank occupied by ARMF-I. It was originally 

installed to complement the capability to measure nuclear characteristics. It was placed into operation 

in December 1962. ARMF-II had a readout system that recorded measurements on data cards, 

resulting in timelier data processing (Stacy, 2000, PDF p. 274). The ARMF-II reactor was modified 

in 1968 and renamed the Coupled Fast Reactivity Measurement Facility (CFRMF). The core was 

modified to produce a region of high-energy neutron flux to provide physics information about the 

behavior of fast (un-moderated) neutrons. Physicists studied differential cross-sections and tested 

calculation methods. The facility contributed to the development of fast-neutron reactors (Stacy 2000, 

PDF p. 275). This reactor was operated until 1991. 

5.1.1.7 Engineering Test Reactor Critical Facility (ERTC)

The Engineering Test Reactor Critical Facility was a full-scale, low-power nuclear copy of the ETR. 

It performed functions similar to those of ARMF and the Advanced Test Reactor Critical Facility 

(ATRC) (discussed below). It was used to determine the nuclear characteristics of fuel and 

experiments planned for irradiation in ETR, and the power distribution effects for a given ETR fuel 

and experiment loading. ETRC enabled operators to predict the nuclear environment when completed 

experiments were removed or new experiments were added. The information was necessary to 

calculate the experiment irradiation and determine core life, control rod withdrawal sequences, 

reactivity worth, and core safety requirements. Mockups of fuel and experiment loadings in ETRC 

were manipulated until a desired power distribution throughout the core was attained, satisfying 

pertinent safety requirements. The ERTC was operated from 1957 through 1982 (Stacy, 2000, PDF p. 

276). Figure 5-13 shows a photo of the ETRC. 
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Figure 5-13: The Engineering Test Reactor Critical Facility (ETRC)

5.1.1.8 Advanced Test Reactor Critical Facility (ATRC)

The ATRC performs functions for the ATR similar to those performed by the ARMF reactors for the 

MTR. ATRC verified for reactor designers the effectiveness of control mechanisms and physicists’ 

predictions of power distribution in the large core of the ATR. Low-power testing in the ATRC 

conserved time so that the large ATR could irradiate experiments at high power levels. The ATRC 

also verified the safety of a proposed experiment before it was placed in the ATR. The ATRC was 

placed into operation in 1964 and is still operational (Stacy, 2000, PDF p. 274). 
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5.1.1.9 Hot Cell Facility

The TRA’s Hot Cell Facility is located in the southwest portion of the ETR building. The Hot Cell 

Facility was first used in 1954 (Stacy, 2000). The building is no longer used for radiological 

operations (DOE, 2009). The facility consisted of three separate cells with a common operating 

corridor. The hot cells were designed and built for the examination of materials exposed to neutron 

bombardment. Processes that were contained in the hot cells included gamma scanning, photography, 

and optical metallography used for handling, photographing, milling, measuring, and weighing 

radioactive samples (DOE, 2009; Stacy, 2000). Each hot cell contained equipment, including lathes, 

power saws, grinders, and welders. There were also stainless steel tables and other fixtures in the 

cells. The hot cells were also used to produce radioisotopes (DOE, 2009). The hot cells were 

constructed as follows.

• Cell #1 (Figure 5-14) was designed for working with high-dose-rate irradiated materials. The 

walls were constructed of nominally 4-ft.-thick high-density concrete, lined on the interior with 

1/4-in. painted carbon steel plate to provide adequate workers shielding. 

• Cell #2 (Figure 5-15) was used for working with lower-dose-rate materials. The walls of this cell 

were formed from ordinary concrete 2-ft-9-in. thick. Floor and lower walls were lined with 1/4-in. 

carbon steel plate. Cell #2 could be divided into two equal-sized subcells by means of a 6-in.-

thick motor-driven steel door which could be slid horizontally on a floor track. Cell #2 and its 

surrounding work area are shown in an overhead photo in Figure 5-16. 

• Hot Cell #3 was completed in 1960 to increase the capacity of the facility; it is located on the west 

end of TRA-632 building. Hot Cell #3 was also used for working high-dose-rate materials. The 

south and west walls were constructed of 5-ft-6-in.-thick ordinary concrete; the remaining walls 

were constructed of 4-ft-thick high-density concrete. The floor and lower walls (to 6 ft above the 

floor) are lined with 1⁄4-in. carbon steel plate. There were four viewing windows on the north wall 

and one viewing window on the east wall. This cell could also be divided into east and west 

subcells using a 6-in.-thick motor-driven steel door as at Cell #2 (DOE, 2009).

The layout of the cells along with offices and other areas is shown in Figure 5-17. 
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Figure 5-14: TRA Hot Cell #1

Figure 5-15: TRA Hot Cell #2
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Figure 5-16: TRA Hot Cell #2

Figure 5-17: TRA Hot Cell Building, Internal Layout
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5.1.1.10 TRA Gamma Facility

The TRA Gamma Facility was located to the south of the original TRA main security gatehouse. The 

facility consisted of a 16-foot-deep canal with cadmium buckets designed to hold spent MTR fuel 

elements. Experimental samples were inserted in sample tubes and then lowered into extremely high 

gamma fields. Sponsors provided a large variety of materials and samples for gamma irradiation, 

including food products and some natural substances such as gold, diamonds, and oil. Irradiated 

samples not radioactive were surveyed thoroughly for external contamination on removal (Stacy, 

2000, PDF p. 126). The Gamma Facility was first operated in 1955; it is no longer operated; the date 

of last use is not determined. 

5.1.1.11 Radiation Measurement Laboratory (RML)

The Radiation Measurement Laboratory, located in the MTR west wing and in use since 1952, was 

previously called the MTR Counting Room. The RML specializes in measuring quantity and quality 

of alpha, beta, gamma, and neutron radiation samples. A variety of counting equipment and 

spectrometers are available in the RML. Over the years, an endless variety of samples have been 

brought for analyses. Some of the detector shielding was made of pre-World War II battleship steel. 

The original equipment has been replaced with more modern equipment. R. L. Heath developed the 

Scintillation Spectroscopy Gamma-Ray Spectrum Catalogues using the capabilities of the RML 

(Conner, 1956; AEC, 1970). 

5.1.1.12 Radiochemistry Laboratories

The Radiochemistry Laboratories, located in the MTR west wing, are used to support RML and 

independent research and development work. Investigators study methods for producing and 

purifying medical radioisotopes as well as the effects of radiation on hazardous waste. Originally, 

Laboratories 109 to 112 were used primarily for chemical analysis of reactor primary systems and 

loop experimental coolants. The predominant radioactivity has been from fission and activation 

products. Figure 5-18 shows a photo of Lab 110. 

The south extension to the MTR Wing was the Alpha Laboratories, which were designed for the safe 

handling of actinides, including Pu-239, Am-241 and Cm-244, and other hazardous alpha-emitters, 

such as Th-232, U-233, and U-235 (Stacy, 2000). Figure 5-19 shows Alpha Lab 127. Figure 5-20 

shows an INL scientist machining thorium in an enclosed glove box in one of the alpha labs in 1964. 

Non-alpha-emitters but with high gamma radiation, such as Pa-233, were sometimes handled in the 

Alpha Laboratories (see Figures 5-20 and 5-21).
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Figure 5-18: Lab 110, MTR Laboratory Wing

Figure 5-19: Alpha Lab 127, MTR Laboratory Wing
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Figure 5-20: Scientist Machining Thorium, Alpha Lab, 1964

Radioisotopes were transferred to this area for analysis or chemical processing in sealed vessels. 

These vessels were packed in cans at the point of origin, which included off-site entities and INL 

entities, such as CPP. Cans were mostly opened in one of the fume hoods where the inner container 

was placed in a clean beaker, still unsealed. The beaker was inserted into an appropriate dry box 

where the container were unsealed, if necessary. There were two types of dry boxes: (1) one type 

handled small quantities of radioisotopes that were not emitting very much gamma radiation into the

glove box; and (2) the other type was used to work with gamma-emitting radioisotopes in large 

quantities. That type of dry box was placed in the large cave facility that used mechanical 

manipulators for handling (Stroschein, 1967).

A stand-alone, highly-shielded hot cell structure was constructed in one of the Alpha Labs in 1961 for 

working with high-activity actinides, other alpha radionuclides, and high gamma-emitting 

radionuclides. This structure was referred to as the “Alpha Cave.” For example, researchers 

separated Am-241 from plutonium in 1962 in the Alpha Cave. A special cask was fabricated to load 

and unload radioactive materials to and from the Alpha Cave (Mahathy, 2015). The alpha cave was 

operated by manipulators, viewports, and remote camera. INL scientists for the first time prepared a 

pure form of the highly-radioactive isotope protactinium-233 in the cave. Figure 5-21 shows 

part of that operation.
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Figure 5-21: Preparation of Pure Pa-233, Alpha Cave

5.1.1.13 HB-4 Crystal Spectrometer

The MTR had several experimental holes placed directly adjacent to the active lattice serving as 

horizontal beam holes (Stroschein, 1967).

The crystal spectrometer occupied the cubicle at beam port HB-4. The spectrometer was used to 

produce a beam of monoenergetic neutrons with wave properties. The atoms in the crystal were 

arranged in an orderly array. When a neutron beam struck the crystal, the atoms of the crystal acted as 

scattering centers. The scattered waves interfered with one another by either reinforcing or cancelling 

each other according to their relative phases (Stroschein, 1967). These beams were used to measure 

relative fission product yields (Phillips, 1962). 

5.1.1.14 Fast (Neutron) Chopper

In order to measure neutron cross-sections of irradiated radionuclides, the Fast Chopper Facility was 

constructed. It used the high-neutron flux available in the MTR for the determination of the 

cross-sections of numerous isotopes of elements. Using beam port HB-6, two shutters controlled the 

reactor beam, directing it through the MTR building to the Fast Chopper. The Fast Chopper building 

was heavily shielded, allowing personnel to work in the vicinity of the chopper for short periods of 

time during reactor operation (Stroschein, 1967). Use of the Fast Chopper was discontinued with the 

use of the MTR. 
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5.1.2 Chemical Processing Plant (CPP)

ATTRIBUTION: Section 5.1.2 was completed by Jason Davis, Oak Ridge Associated Universities. All 

conclusions drawn from the data regarding the Chemical Processing Plant were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the Chemical Processing Plant are explained in the associated text.

Construction of the Fuel Reprocessing Complex at the Chemical Processing Plant started in 1950 with 

the Bechtel Corporation serving as construction contractor and American Cyanamid Company as 

operating contractor. After three years of construction activity and extensive testing, the plant was 

ready to handle its first load of irradiated fuel. Phillips Petroleum took over as operating contractor in 

1953, shortly after fuel-reprocessing activities became routine. 

The core of the original Fuel Reprocessing Complex consisted of seven structures, a pair of deep 

wells, and a power substation. The complex expanded and, at its peak, consisted of more than 100 

facilities and support buildings located on approximately 200 acres. Table 5-2 lists the CPP facilities 

of concern that are described in the following subsections. Figure 5-22 shows a map of the CPP 

facilities.

Table 5-2: INL Chemical Processing Plant (CPP) Processing and Support Facilities

Building Number Facility Name First Used Last Used

CPP-601 Main Processing Building 1953 1992

CPP-602 Laboratory 1952 2009

CPP-603 Fuel Storage Building 1953 Present

CPP-604/605 Waste Disposal Building and Rare Gas Plant 1953 ND

CPP-620 Chemical Engineering Laboratory 1968 ND

CPP-627 Remote Analytical Facility and Multi-Curie Cell 1955 1997

CPP-630 Mass Spectrometry Facility 1956 2009

CPP-631 RaLa Off-Gas Filter Building 1957 1963

CPP-633 Waste Calcining Facility 1963 1981

CPP-637 Process Improvement Facility 1959 ND

CPP-640 Hot Pilot Plant/Headend Process Plant 1963 1981

CPP-709 Eastside Service Waste (ESSW) Monitoring Station 1952 1990

CPP-734 Westside Service Waste (WSSW) Monitoring Station 1960 1989

CPP-1781 High Level Liquid Waste Tank Farm 1951 Present

Multiple Calcined Solids Storage Facilities (CSSF) 1959-1984(a) Present

ND = Date of last use not determined

(a) Calcined Solids Storage Facilities were built as needed and put into service as they were completed. 

Following the facility descriptions there are process overviews of these operations of concern:

• Overview: Fuel Reprocessing Operations 

• Overview: Isotope Recovery

• Overview: Waste Calcination
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Figure 5-22: Map of CPP Facilities
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5.1.2.1 CPP-601, Main Processing Building

The main processing building, CPP-601, which housed the bulk of the processing equipment and 

controls, was a simple-looking structure with extensive below-ground facilities (see Figures 5-23 and 

5-24). It had two levels. The lower level had four stories and was constructed of steel-reinforced 

concrete. The upper, above-ground level was one story constructed of Transite and structural steel.

Figure 5-23: Locations of CPP-601, CPP-602, CPP-630, and CPP-640

The top story of the building was an un-partitioned space for storage, make-up of chemical solutions, 

and transfer of fuel elements to the processing equipment below; it was known as the Process Makeup 

(PM) Area. Below the Process Makeup Area, CPP-601 was divided into a number of corridors and 24 

cells. 

The 24 cells were arranged in two parallel rows with Operating, Service, and Access corridors 

extending down the middle. All the equipment for the actual processing of spent fuel elements was 

contained in these two rows of shielded cells, each uniquely identified by an alphabetic designator. 

(Rooms housing vacuum equipment, pumps, and tanks were also called cells, although they were 

usually not given alphabetic designations.) Concrete walls ranging from two- to five-feet thick

provided shielding. At the Operating Corridor level, these thick walls were honeycombed with offset 

pipe sleeves for entrance and exit of utility, process, and instrumentation lines. Most of the equipment 

within the cells was constructed of stainless steel or other acid-resistant materials. Only one cell 

(L-Cell) was equipped with a viewing window; for the other cells, the equipment was not visible 

during operations. 
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Figure 5-24: Perspective Drawing of CPP-601, Main Processing Building

Below are descriptions of the CPP-601 cells: 

• A-Cell is located in the western bank of processing cells. The EBR-I batch dissolvers, various 

tanks, and an off-gas condenser installed in the cell were designed specifically to process 

nearly-pure uranium that had once been clad in stainless steel to provide power for EBR-I. The

material was used only once, in 1954. In the 1960s, all of the equipment in A-Cell was viewed as 

expendable, to be replaced as necessary to increase the fuel-processing capability of the CPP

(Stacy, 2006).

• B-Cell is located directly below A-Cell in the southwest corner of CPP-601. B-Cell originally 

held sixteen tall slender tanks, sized and spaced to provide critically-safe storage of concentrated

solutions of highly-enriched uranium from various dissolvers in the plant. While the tanks 

themselves were geometrically safe, in 1959 a small amount of solution siphoned from one of the 

tanks into the process equipment waste system, resulting in a criticality. After decontamination, 

the cell and tanks were used once again for interim storage of various chemical solutions, 

including a mixture of neptunium and plutonium (from 1972 to 1977) (Stacy, 2006).
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• C-Cell is also located in the west bank of processing cells adjacent to A and B Cells. It is 

essentially identical to its neighbor, D-Cell, and often worked in tandem with it to dissolve 

aluminum-alloyed fuels. In 1953, the batch dissolvers in C-Cell and D-Cell processed the first 

batch of hot fuel introduced into CPP-601. Many more batches of aluminum-alloyed fuel followed 

this pioneering run (Stacy, 2006).

• D-Cell is a mirror image of C-Cell. Both were originally designed for batch dissolution of fuels 

with aluminum cladding and each held a complete system for this so-called “headend” process (for 

removing fuel cladding) (Stacy, 2006).

• E-Cell housed the primary equipment for the original batch zirconium dissolver, made of Monel to 

resist the corrosiveness of the hydrofluoric acid used to dissolve the zirconium. Another vessel in 

E-Cell, made of Carpenter-20 steel, was used for the early two-step sulfuric/nitric acid dissolution 

of fuels clad in stainless steel. The zirconium dissolution equipment was modified several times 

during its processing history; first, to accommodate semi-continuous processing, and later, to 

support co-processing with solutions from the aluminum “headend” process. Most of the fuels 

processed through the E-Cell dissolvers originated with the U.S. Navy, including cores from the 

Nautilus and Sea Wolf submarines (Stacy, 2006).

• F-Cell floors and walls were lined with stainless steel. After being dissolved in one of two vessels 

in E-Cell, Navy fuels were sent to geometrically-safe equipment in F-Cell for the first cycle of 

solvent extraction using tributyl phosphate (TBP). This extraction equipment was last used in 

1965 and was largely replaced by a new technology, electrolytic dissolution, housed in CPP-640. 

After decontamination, F-Cell was modified to serve as a feed clarification system for graphite 

and aluminum dissolver products. Centrifuges installed in F-Cell at this time were instrumental in 

removing the solids from these solutions to facilitate uranium extraction (Stacy, 2006).

• G-Cell was one of the busiest cells in the entire plant. This cell housed two continuous dissolvers 

for aluminum-alloyed fuels plus the large scale first-cycle extraction columns that accepted 

solutions from many different dissolution systems in the plant. The dissolvers first operated in 

1957 and were used again and again until 1986. The G-111 column, which completed the initial 

cycle of solvent extraction of uranium from a variety of dissolver products without fail for more 

than twenty years, has been called the “workhorse” of the plant. In addition to myriad runs with 

dissolved aluminum alloys, in 1969, G-111 would process the first dissolver product from coprocessing of zirconium and aluminum. Years later, in 1987 and 1988, the last batch of fuel to be 

dissolved in CPP-601 would pass through. G-Cell also contained several holding tanks for highlevel liquid wastes produced during first-cycle extraction. These raffinates were extremely 

contaminated both physically and radioactively and could be jetted directly to large 300,000 gallon 

cooled storage tanks in the CPP Tank Farm (Stacy, 2006).

• H-Cell is located next to G-Cell on the west side of CPP-601. These cells were identical in size 

and also worked in tandem to process aluminum-alloyed fuels and a first-cycle extraction system 

for all types of dissolver products from the plant. Generally, H-Cell equipment received the 

stream of uranium-bearing liquid isolated in G-Cell, processed and purified it further, and sent the 

resulting high-level liquid waste back to G-Cell for temporary holding. The columns, tanks, and 

evaporators that processed uranium-bearing solutions in H-Cell are geometrically-unsafe under 
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some circumstances so operators had to be very mindful of uranium concentrations as solutions 

were being processed there. Even with strict controls in place, two criticality incidents occurred in 

the cell during its processing history. The first, in 1961, was centered in an evaporator; the last, in 

1978, was in a scrub column. Modifications to the processing equipment, particularly the 

evaporators, and tighter administrative controls on additions to the process stream, were designed 

to prevent these types of problems in the future (Stacy, 2006).

• NOTE: No cell within CPP-601 was designated as I-Cell because of the potential for confusion 

with the number one (1) (Stacy, 2006).

• J-Cell was always devoted to uranium salvage operations. However, it did go through three main 

changes in configuration during its processing history. Initially, the cell contained a double bank 

of tanks that continually created problems because they were of insufficient capacity for plant 

operations. This problem was addressed by installation of a system of fixed nuclear poisons that 

was intended to increase the capacity of the salvage system without creating criticality problems. 

These controls did not work as expected and criticality problems did arise. As a result, in its final 

configuration, the cell was restricted to processing of relatively low-level and low-quantity 

solutions from the plant’s process equipment waste system (Stacy, 2006).

• K-Cell was devoted to recycling operations. Solvents, including both hexone and TBP, were sent 

to scrubbers and evaporators in K-Cell for decontamination and purification. Uranium separated 

during this process went back to one of the uranium salvage operations (C, J, and L-Cells) for 

further purification while the clean solvents were temporarily stored for use at a later time (Stacy, 

2006).

• L-Cell arguably became the most complex cell in CPP-601, even though it started out empty and 

was intended as a spare for future processing activities. Although it started out without a mission, 

L-Cell did not remain idle for long. Changes began to appear in 1954: a shielded viewing window, 

remote handling manipulators, a periscope, and intercom system were installed that would allow 

an operator to see and hear the work going on behind six feet of concrete shielding (see Figure 

5-25). A wide variety of equipment also began to appear inside the cell. The equipment included 

an elevator, two large centrifuges, a dissolver and its charging chute, an off-gas scrubber, and 

process tanks of many sizes. The set-up that emerged from these changes was designed to isolate 

an intensively radioactive and short-lived fission product, lanthanum-140 (radioactive daughter of 

barium-140), from fuel elements fresh from the MTR. By the time most MTR fuel elements 

reached the CPP processing cells (after at least 120 days of cooling), all of the lanthanum had 

decayed. However, the fuel brought to L-Cell was new enough that useable quantities of it were 

still present. When the first usable amounts of this isotope were isolated in L-Cell in 1957, it 

became known as the RaLa Cell, which was short for the radioactive lanthanum produced there 

and quickly shipped to Oak Ridge for research. The process continued until 1963. L-Cell 

remained idle for many years until 1984, when a new critically-safe uranium salvage system was 

installed to process solutions that were known or suspected to contain recoverable amounts of 

uranium (Stacy, 2006).
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Figure 5-25: Operating Face of L-Cell

• M-Cell was also originally built as a spare in the eastern row of cells. The principal function of 

M-Cell was to provide temporary storage and sampling capability for concentrated uranium 

solutions flowing from the first-, second-, and third-cycle solvent extraction operations. Four 

tanks and a variety of pumps complete these tasks and supply the necessary measurements of 

accountability for the plant. Due to the potential need for maintenance, the pumps are shielded 

from the vessels by an internal concrete wall that is 1.5 feet thick by 15.5 feet tall (Stacy, 2006).

• N-Cell housed six critically-safe storage tanks that were used to temporarily store aqueous 

uranium solutions from first-cycle extraction that were awaiting processing through second and 

third cycles of the process. One tank was also used to store a specialized neptunium concentrate 

until enough had accumulated for a unique processing run. Although the tanks were critically safe 

by geometry, a criticality event in B-Cell in 1959 (also used to store intercycle solutions) made it 

clear that there was potential for inadvertent damage of concentrated uranium solution to the 

process equipment waste system and a corresponding vulnerability for criticality. To prevent this, 

CPP engineers adapted a standard device for criticality control, Raschig Rings. (Raschig Rings 

are pieces of tube (usually ceramic or metal, equal in length and diameter) used in large numbers 

as a packed bed within columns for distillations and other chemical engineering processes. They 

provide a large surface area for interaction between liquid and gas or vapor.) These rings were 

usually installed in the processing columns. Instead, in this case, they were installed on the floor 

of the N-Cell, where they could keep concentrations of uranium down to acceptable levels. 

Although the original glass Raschig Rings proved to be too delicate for direct maintenance, a 

stainless steel mesh installed over the top of them solved that problem and allowed the cell to 

operate for many years (Stacy, 2006).
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• NOTE: To prevent confusion with the number zero (0), no operating cell within the original floor 

plant of CPP-601 was designated as O-Cell. Later however, a pump room located to the westnorthwest of N-Cell began to carry the name. The pumps installed in O-Cell moved solutions 

between the first-, second-, and third-cycle extraction equipment (Stacy, 2006).

• P-Cell has always been used for solvent extraction of highly-enriched uranium and often worked 

in tandem with similar systems installed in Q and R/S Cells. It was well-suited as a home for the 

tall equipment that purified the uranium at this stage of the process; six columns used for 

extraction, stripping, and evaporation occupied more than half of the long narrow space. Most of 

the time, P-Cell equipment was used to process second-cycle solutions using hexone as an organic 

solvent. For a short time early in its history, it was employed for first-cycle extraction. In 1972, a 

special solution of concentrated neptunium and plutonium that had been collected from seven 

years of fuel-reprocessing activities was also processed through the extraction equipment in 

P-Cell. Then, after thorough decontamination, it was ready to once again handle concentrated 

uranium (Stacy, 2006).

• Q-Cell was designed to take second-cycle uranium solutions from P-Cell and process them 

through a third, and usually final, cycle of solvent extraction. Hexone was also used as a solvent 

in Q-Cell’s tall columns that were used for extraction, scrubbing, stripping, and evaporation. 

Sometimes, however, Q-Cell was used to process solutions directly from first-cycle extraction 

(Stacy, 2006).

• R-Cell does not exist as a separate entity from S-Cell since there are no physical separations 

between them. Originally, R/S-Cell contained extraction equipment for a fourth and final cycle of 

solvent extraction and uranium purification. However, after continuous dissolvers were developed 

and put into routine use at the plant, uranium product from the third-cycle extraction equipment in 

Q-Cell rarely failed to meet specifications and R/S-Cell extraction columns soon became obsolete. 

Once cleared, the R/S-Cell space was used again though, particularly in the monitoring of other 

plant process streams. In 1979, a tank and a system were installed for measuring the density of 

concentrated uranium solutions. It was hoped that this density monitoring system would help with 

criticality problems in the uranium salvage system installed in J-Cell. Then, after the J-Cell 

uranium salvage system was replaced, the density monitoring tank in S-Cell was modified to be 

used as a decanter to isolate TBP and prevent it from entering the uranium salvage system 

installed in C and L-Cells (Stacy, 2006).

• S-Cell contains the bottom half of all of the equipment installed in R-Cell. The two cells are one 

in the same in terms of physical space. Thus, like R-Cell, S-Cell also went through three different 

iterations during its processing history. Originally, it held fourth-cycle extraction equipment. In 

1979, a density monitoring tank was installed to help with criticality problems in J-Cell and, in 

1986, a decanter was installed to remove organic solvent from the uranium process stream (Stacy, 

2006).

• T-Cell was used, from the beginning, as a reservoir for cold hexone that was received from K-Cell 

or added fresh from the Process Makeup floor below. The solvent was stored in T-Cell tanks, 

ready to be pumped to P- or Q-Cell as needed for uranium extraction operations. Pumps and 

blowers facilitated this transfer (Stacy, 2006). 
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• U-Cell lies directly beneath T-Cell in the east cell bank. It is one of the two cells in CPP-601 (Uand Y-Cell) that were devoted to the collection of aqueous waste streams from the remainder of 

the plant. Originally, U-Cell collected raffinates from first- and second-cycle extraction 

operations, sampled them to verify that they were critically safe, processed them through 

evaporators to reduce their volumes, and then sent them off to the Tank Farm for long-term 

storage. However, the volume reduction achieved through waste calcining would soon make the 

evaporators unnecessary and they were abandoned in place. Additionally, safety upgrades during 

the 1980s led to a reconfiguration of U- and Y-Cells to combine all sources of second- and 

third-cycle wastes into a single stream that could be independently sampled for criticality at two 

different points (Stacy, 2006).

• V-Cell was held as a spare throughout the history of CPP-601 fuel reprocessing. No processing 

equipment was ever installed. Instead, it served as an office space for health physics personnel 

working in CPP-601 (Stacy, 2006). 

• W-Cell contained tanks and equipment that captured small amounts of the organic solvent, 

hexone, that lingered in the raffinate streams from first-, second-, and third-cycle extraction. 

Eventually, lines were rerouted so that all hexone collection was combined in this cell (Stacy, 

2006).

• X-Cell occupies the northeastern corner of CPP-601. Like V-Cell, X-Cell never held equipment 

for fuel processing. However, for a short time it was used as a laboratory for storage and analysis 

of highly radioactive samples, for the dilution of samples before transport to other laboratories, 

and for decontamination of the equipment used in sample preparation. These activities were 

carried out in six shielded areas constructed from lead-filled steel forms and lead bricks within the 

cell. A monorail and crane transported samples from one area to another or they were manually 

carried. From the beginning, this lab was plagued with contamination problems; construction of 

the Remote Analytical Facility (CPP-627) in 1957 made it obsolete. Pumps and blowers added 

later changed the focus of the cell to solvent recovery and storage (Stacy, 2006). 

• Y-Cell was the second of two cells used to collect liquid raffinates from process streams 

throughout the entire fuel reprocessing plant. Initially, Y-Cell collected some second-cycle 

streams and all third-cycle streams, sampled them for criticality, and then sent them through 

evaporators to reduce their volume prior to long-term storage in the Tank Farm. However, 

eventually all second- and third-cycle wastes would flow through the equipment in Y-Cell for 

sampling and then transfer to the Tank Farm. The dissolvers were abandoned after being made 

obsolete by the waste calcining facility. The solvent recovery system was also removed when the 

liquid waste collection process was consolidated in W-Cell (Stacy, 2006).

• Z-Cell was the last stop for CPP-601’s final product, concentrated liquified uranyl nitrate. For a 

time, the liquid was stored and packaged for shipment in L-10 bottles at a facility set up within 

Z-Cell. However, after 1969, the liquid was converted to a solid (uranium trioxide) that was easier 

to handle and ship. The denitrator used in this process, located in the CPP-602 Laboratory 

building next door, was developed using local expertise acquired during process development for 

the CPP’s waste calciner (Stacy, 2006).
 

 

 

 

 

 

 

 

 

 

 

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Table 5-3 provides process descriptions and source documents for the cells in CPP-601, the Main 

Processing Building. 

Table 5-3: CPP-601, Main Processing Building Cells 

Cell Process Description Reference 

A EBR feed preparation 

Irvine, 1954 B EBR feed storage 

C NP-MTR feed make-up 

D NP-MTR feed make-up 

E SIW feed preparation 

Stacy, 2006 F First-cycle extraction 

G MTR feed preparation 

H MTR first-cycle extraction 

J Hot salvage Irvine, 1954 K Solvent recovery 

L RaLa process cell Stacy, 2006 M Sampling 

N NP-MTR feed storage Irvine, 1954 

O Pumping Stacy, 2006 

P First/second-cycle extraction 

Q Second/third-cycle extraction Irvine, 1954 

R Product transfer cell 

S Third-cycle extraction Stacy, 2006 

T Solvent pump room 

Irvine, 1954 

U First-cycle aqueous raffinate 

V Equipment decontamination 

W First-cycle solvent raffinate 

X Sample dilution 

Y Second/third-cycle aqueous and solvent raffinate 

Z Product storage 

Equipment in the cells could be reached and moved with large cranes through hatch openings in the 

ceilings, but plant workers could only enter through doors installed in the Access Corridor three floors 

below ground. 

Each of the shielded cells housed equipment necessary for performing various steps in the 

reprocessing cycle. Most of the process equipment within the cells was controlled from an Operating 

Corridor (PO) that ran the length of the building between the two rows of cells in the first story below 

ground. A double row of instrument panels arranged back-to-back was installed in the center of this 

gallery. From here, operators could monitor the workings of the cells below and control the flow of 

solutions into and out of the processing cells. Diagrams and chemical flowsheets that illustrated all of 

the main processes conducted in the plant were posted there to guide the operators when necessary. 

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Service lines for water, steam, and condensate also entered the cells at the Operating Corridor level. 

Directly beneath the centralized Operating Corridor were the Service and Access Corridors (second 

and third stories, respectively). Sampling corridors ran along the outside of each cell row along with 

cell exhaust ventilation ducts and off-gas treatment systems.

The Sampling Corridors in CPP-601 ran along the outside of each bank of processing cells on the 

same level as the Operating Corridor (ground level). Because the dissolution and extraction processes 

relied on specific chemical inputs, sampling was an important part of process control for the plant, 

ensuring that the end product was as pure as possible and preventing formation of troublesome solids 

or chemical ions that might clog the equipment or otherwise cause a plant shutdown. Sampling was 

also very important for accurate accounting of the fissionable material present in the plant at any one 

time. Once drawn, the sample bottles were placed in lead carriers and manually transported to the 

Remote Analytical Facility, CPP-627, for analysis. 

The Service Corridor (PT) that runs the length of CPP-601, directly beneath the Operating Corridor, is 

the first basement story of the structure. Shielding precautions were taken because nearly all of the 

pipes that carried “hot” radioactive solutions between processing cells were located here. Ventilation 

tunnels also ran the length of this level along the outside of the cell rows. Air from the processing 

cells and Sample Corridor was exhausted through these tunnels and then ultimately delivered to the 

main stack for discharge. The ventilation system provided twenty air changes per hour to the process 

cells, primarily to sweep out any potentially-explosive solvent vapors that might have accumulated 

there. Negative pressure, maintained in all of the processing areas, effectively ensured that 

radioactive contamination did not spread into the ventilation system. Vapors directly from the 

processing equipment, which could have contained radioactivity, were collected in special piping 

systems in the ventilation tunnels and then delivered to CPP-604 for filtration through off-gas 

systems.

The Access Corridor (PA), located in the third level below ground in CPP-601, contained the 

doorways that workers used to gain entry to the interiors of all of the processing cells. This corridor 

also served as a fresh-air intake for the plant. Fresh air was pulled into the processing cells through 

louvers in the doors. When these doors were opened for entry of decontamination, maintenance, 

and/or construction workers, the negative pressure maintained within was reduced. Temporary tents 

were often placed at the entrance to cells that were to be opened to help control airflow and reduce the 

spread of contamination outside the shielded cell walls. Labyrinth corridors were also constructed at 

the entrance to each cell and around some of the pumps and other equipment located outside of the 

cells to prevent any direct shine of radiation out into the Access Corridor. 

5.1.2.2 CPP-602 Laboratory

The CPP-602 Laboratory was located north of and adjacent to the Fuel Process Building (CPP-601), 

with which it shares a common wall and utilities. The interior of the CPP-602 Laboratory was divided 

into offices, laboratories, and other areas with partitions made of steel, wood and gypsum board, or 

cinder block. 

The exhaust air from occupied areas, hoods, glove boxes, and the hot cell was drawn by blowers 

through roughing and high-efficiency particulate air (HEPA) filters for venting. All air was exhausted 

through HEPA filter systems in CPP-602 Room 330, on the roof of the CPP-630, or vented via the 
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CPP-601 vent tunnel to the Atmospheric Protection System (APS) in CPP-649 via the CPP-601 east 

vent tunnel. The APS vents to the atmosphere via the Main Stack.

Most of the air in the basement was exhausted via the service corridor of CPP-601 or went to the east 

vent tunnel via the hot cell, glove boxes, and hoods of Laboratory 103A and the hood of Laboratory 

121B. The northeast hood in Laboratory 103A, the hoods and glove boxes in Laboratory 103B, the 

hood and denitrator product repackaging glove box in Room 109, and the hoods and glove box in 

Room 325 on the second floor were exhausted through HEPA filters on the roof of CPP-630. All of 

the air on the first-floor laboratories, Laboratories 315 and 327 on the second floor, and the product 

handling Room 102 was exhausted through HEPA filter banks in Room 330 to the roof of the 

CPP-602 Laboratory (Grigg, 2008). 

5.1.2.3 CPP-603, Fuel Storage Building

The safety and security concerns that dictated placement of the Fuel Storage Building, CPP-603, 

one-third mile away from the Main Processing Building and other analytical facilities probably also 

entered into the decision to install systems that made the building nearly self-contained within the 

remainder of CPP. 

The building was designed to receive fuel elements in a main crane bay via a 15-ton crane. A long 

transfer canal paralleled the crane bay; branching off from there were two basins where the fuel was 

stored. The two basins provided storage for about 1,000 buckets that could hold as many as 4,000 

separate fuel elements, depending on overall size and criticality considerations. At least 15 feet of 

water covered all radioactive materials that were handled outside of the lead casks used for 

transportation.

In 1957 and 1958, CPP-603 was expanded in response to planned shipments of fuel elements from the 

Savannah River Site. The new “Fuel Element Cutting Facility” added to CPP-603 at that time housed 

a large crane bay serviced by two cranes, a new storage basin, and a hot cell. Large fuel elements in 

casks weighing up to 75 tons were received in this facility via truck or rail and then unloaded into the 

transfer basin where they could be removed. Once the fuel elements were free of their shielded casks, 

operators used a 3,000-pound transfer crane, complete with a riding car, to place them in cadmiumpoisoned storage racks which could then be moved into the new storage basin. Oversized pieces were 

delivered to the new hot cell where operators used hydraulically-operated saws and other equipment to 

mechanically alter their shapes and make them suitable for storage and reprocessing at CPP. 

In 1974, as CPP began to experiment with, and ultimately begin production-level processing of, 

graphite-based fuel elements, facilities for dry storage were added to CPP-603. These additions were 

necessary because the graphite fuel elements reacted violently if exposed to water, and thus could not 

be stored in any of the existing underwater basins. Called the “Irradiated Fuels Storage Facility,” this 

was the first dry-storage facility designed, built, and actively used in the U.S. Initially, the facility 

consisted of forty-seven underground vaults, each measuring sixty square feet, lined with stainless 

steel, and capped with removable heavy concrete covers. In 1975, another building was added to 

handle more shipments of graphite fuels. 
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5.1.2.4 CPP-604/605, Waste Disposal Building and Rare Gas Plant

Both liquid and gaseous wastes were processed in the Waste Disposal Building (CPP-604). Here, 

liquid wastes were evaporated in order to reduce their volume, then sent for storage in the tank 

farms. However, the gaseous wastes were processed in a portion of the building designated as the 

off-gas area (CPP-605). Here, isotopes of krypton and xenon were recovered at the CPP from offgas produced in the aluminum-clad fuel dissolver (Irvine, 1954). 

5.1.2.5 CPP-620, Chemical Engineering / High-Bay Laboratory

The High-Bay Laboratory was built in 1968 as a chemical engineering laboratory facility where 

versatile equipment was located. The facility was used primarily for non-radioactive testing of plant 

processes and for the development and improvement of new processes. However, some projects were 

carried out with un-irradiated uranium. The facility was designed to house tall equipment and to allow 

for embodied experimental equipment to be installed and removed in prefabricated modules. 

The major equipment in CPP-620 consisted of the equipment housed in experimental modules and the 

chemical makeup equipment that was fixed in position. Each module was a structural-steel 

framework containing the equipment associated with the particular experiment. Vessels, pumps, 

stirrers, piping, and valving associated with chemical makeup were located in an area between the 

south office and the south wall of the building. Instrumentation, control panels, and electrical 

equipment panels associated with the various experimental modules were located on the Mezzanine 

level. The building also had three offices arranged down the center surrounded by the experimental 

areas. The exterior tops of the offices constituted a control area at the Mezzanine level which 

connected with the experimental modules by catwalks.

No fixed radiation instrumentation was installed in the High-Bay Laboratory. A portable Continuous 

Air Monitor (CAM) was located against the north wall of the west experimental area to provide 

radiation monitoring coverage during those infrequent times when radioactive materials were used. 

This monitor was equipped with a strip chart and meter to give both a visual and a permanent record 

of air activity. A bell alarmed to warn personnel when air activity reached a preset limit (CPP Safety 

Review, 1974).

5.1.2.6 CPP-627, Remote Analytical Facility

The Remote Analytical Facility (RAF), CPP-627, was home to a variety of customized dissolution 

processes, but it also filled an important analytical role for the plant. The building was constructed in 

1955 to house analytical, experimental, and decontamination facilities. The analytical portion of the 

building, located on the ground floor within the northern third of the building, consisted of thirty-two 

shielded glove boxes, each measuring about one meter square, and arranged in a row much like the 

operating corridor in CPP-601 (see Figure 5-26). Viewing windows and manipulators in these boxes 

allowed for remote sample preparation and analysis.

The northern third of the building housed radiochemical analytical facilities. The RAF, consisting of 

two lines of shielded cells for remote sample preparation and analysis, was on the ground floor. The 

Old Shift Laboratory (OSL, Room 201) on the second floor provided bench and hood space for 

chemical analysis of nuclear reactor fuel samples of low-to-moderate activity. The OSL operated in 
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conjunction with the RAF to supply 24-hour analytical services in support of CPP-601 and calciner 

operations. Analytical services were provided round-the-clock to plant operations (INEEL, 2004). 

The middle third of the building was a high bay decontamination laboratory, providing space for water 

and chemical cleaning of contaminated equipment from all over the NRTS. 

Figure 5-26: Shielded Glove Boxes in the Remote Analytical Facility

Like most of the facilities at CPP, CPP-627 was in constant evolution. In 1956, only six short months 

after the first hot samples were introduced for analysis, modifications were made in response to 

feedback from the scientists who had been using the new equipment. As a result of these analysts’ 

concerns, remote mechanical arms and manipulators replaced the more clumsy remote switches and 

gears originally installed in the glove boxes. 

The southern third of CPP-627 contained two experimental facilities, the Hot Chemistry Lab (HCL, 

Room 104) and Multi-Curie Cell (MCC, Rooms 104 and 106). Both were used for small-scale 

custom dissolution processes and other hot analytical work (INL, 2006). A large walk-in hood 

installed in the HCL was home to some of CPP’s custom dissolution equipment and the remainder 

was installed in the MCC. This cell was shielded to the same degree as the main processing cells in 

CPP-601, enabling researchers to mock up near-production-level processes in their experiments. In 

1972, a solution containing 5,412 grams of neptunium and 544 grams of plutonium was pumped from 

the CPP-601 processing equipment to a temporary packaging facility in the CPP-627 Multi-Curie 

Cell. There it was bottled, packaged, and shipped in specialized containers to the Savannah River Site 

for further purification and ultimate conversion to Pu-238. 
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All work in the MCC was conducted remotely with master-slave manipulators for light work, and a 

2,000-pound hoist for removing cask lids and other heavy work. A 15-ton cask dolly delivered the 

casks of fuel to the cell through a massive, 18 in.-thick, lead-filled door that weighed 20 tons. 

5.1.2.7 CPP-630, Mass Spectrometry Facility

CPP-630 was located adjacent to the east side of CPP-602, with which it shared a common wall. The 

internal laboratory walls were of steel construction and the floor was a base of poured concrete on 

earth, with no buried utilities. The Mass Spectrometry Facility was part of the Radioanalytical 

Chemistry Department, and the majority of the samples processed in this lab were final-product 

samples containing mostly uranium (CPP Progress Report, Sep1955). Laboratory air from the various 

laboratories, hoods, and glove boxes was exhausted through HEPA filters and blowers located on the 

roof of CPP-630. 

5.1.2.8 CPP-631, RaLa Off-Gas Filter Building

The RaLa off-gas system was designed and installed to capture the gases generated during the 

operation of the RaLa process system in L-cell of CPP-601. The RaLa process posed two problems: 

(1) the activity hazard due to the liberation of radioiodine and radioxenon; and (2) the build-up of 

explosive hydrogen gas. The radioactivity of the liberated gases was too high to allow for venting to 

the atmosphere, and the concentration of hydrogen made mechanical compression and storage of the 

gases dangerous. 

Originally, a caustic scrubber was installed in L-cell for processing these gases. However, there was 

no means of storing the gases for decay in the volume and composition encountered. A 10,000-cubicfoot tank was installed in July 1957 to hold the gases that made it through the scrubber. It was quickly 

noted, however, that radioiodine could escape from the system without passing through the scrubber. 

The solution to the problem was the installation of a separate RaLa off-gas handling system with the 

storage tank, blowers and jets for venting, and carbon beds for iodine removal (Smith, 1981). 

The RaLa Off-Gas Filter Building, also known as the RaLa Off-Gas Cell, was an underground 

concrete structure constructed in the earthen berms outside the southeast corner of CPP-601. The 

building contained a centrifugal blower, carbon beds enclosed by lead shielding walls, filters, a 

hydrogen analyzer, and all of the utility connections, ventilation, heating, and instrumentation 

necessary for equipment operation. Piping between CPP-601 and CPP-631 ran through an 

underground concrete tunnel. This tunnel also served as a ventilation duct into the vent corridor of 

CPP-601, providing a slight negative pressure inside the Off-Gas Building (Moser, 1985).

5.1.2.9 CPP-633, Waste Calcining Facility

Construction on the Waste Calcining Facility (WCF) began in 1958 and was completed in 1961. The 

WCF building was built east of CPP-601 and south of the tank farm. The building contained 

everything required for the calcining process except for the tanks that stored fuel oil and the bins that 

would store the calcined product.

The sub-grade processing cells were arranged in two parallel banks with a shielded corridor between 

them. This corridor had operation control panels, monitors, and switches. The calciner vessel had 
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elaborate piping systems that fed it with heat, liquid waste, and hot fluidizing air, and carried away its 

byproducts. Off-gases went to the CPP stack; the calcine went to storage bins (INEEL, 1998).

The main floor level was built four feet above grade and contained five rooms. A Health and Safety 

Office included a small shielded cubicle for sampling the gases that were leaving the building. A 

Switch Gear Room contained the electrical power controls for the facility. A Locker Room for 

personnel contained bathrooms and hand and foot counters. The Decontamination Room contained 

tanks filled with the chemicals that made up the solutions added to the waste feed, or that flushed the 

operating cells after a campaign. A hopper containing the calciner start-up material also was located 

here, along with compressed air equipment, steam manifolds, and solution-transfer pumps. The fifth 

room was the Heating and Ventilating Room. Outside air entered the supply unit, was filtered, heated, 

washed, and reheated before it was sent to the rest of the main floor or to the operating areas below. 

A special equipment room for handling NaK (sodium-potassium alloy) was located three and a half 

feet below the main floor and just above grade level. The location of this room prevented leakage into 

the other rooms or cells of the building because it was actually a separate building connected to the 

main level by a hallway, but sealed by two doors. The room contained the NaK furnace, tanks, 

fluidizing air blowers, pumps, indicators, and a helium manifold system. To accommodate the 

furnace, its ceiling was 30 feet high. 

A Filter Removal Room was approximately at grade level, also. Off-gases from the calciner were 

passed through filters to remove fine radioactive and other particles. The filters were in three cells 

just below the Filter Removal Room and, when spent, had to be replaced with fresh filter units. 

Operators worked outside this room and managed the filter-replacement procedure remotely. They 

stood in the filter-removal corridor behind a shielded glass viewing window (40 inches thick) to 

operate the crane and other equipment. A 20-ton bridge crane lifted the four-foot thick concrete cell 

covers, helped place the spent filters in special shielded transport casks, and then placed the casks for 

pick-up by "straddle carriers." The carriers took them to a waste storage facility elsewhere at NRTS. 

The operating corridor rested below grade and could be accessed via a stairway from the main floor. 

The main instrument panel was centered in this corridor, while service piping for the cells ran along 

the sides. Also located at this level was the Sample Room, from which operators could collect a 

sample of either the liquid waste or the calcine for transport to a hot cell or laboratory for analysis. 

The access corridor provided access to the process cells and related equipment located directly below 

the operating corridor. On each side were the hot process cells. Technicians could access the calciner 

and other cells by going through a labyrinth of concrete shields. A Waste Hold Tank Cell was at the 

east end of the corridor. The tanks held the liquid waste solutions awaiting feed into the calciner. The 

solutions (including spent decontamination chemicals) rested here until the next campaign, when they 

became part of the feed into the calciner.

Next to this was the large Calciner Cell, which contained the calciner vessel itself, the NaK heat 

exchanger, the calciner cyclone, feed-metering pot, and other related equipment. Maintenance 

technicians entered this cell through a shielding labyrinth. Next to the Calciner Cell, the off-gas cell 

contained quenching tanks, pumps, venturi scrubbers, and other equipment related to the handling of 

the gases liberated by the calciner. Other cells included a hot sump tank cell (situated beneath the 
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access corridor floor), adsorber manifold cell, two adsorber cells for ruthenium-106, and a filter piping 

tunnel (INEEL, 1998).

For maintenance during shutdown periods, crews used remotely-controlled cleansing sprays to flush 

each operating cell. Equipment pertinent to the operation of each cell was located just outside the cell, 

typically encased in shielded lead or steel cubicles. Crews used ladders, platforms, and removable 

hatches to gain access to the equipment. Most equipment was made of stainless steel to resist the 

corrosive attacks of decontaminating solutions and the acids of the radioactive feed itself. 

5.1.2.10 CPP-637, Process Improvement Facility / Low-Bay Laboratory

Process support and pilot plant studies were carried out in the labs of the Process Improvement 

Facility (PIF). Also known as the Low-Bay Laboratory, the PIF was a chemical engineering 

laboratory facility built in 1959 with functions very similar to those performed in CPP-620. 

Equipment was located in this facility primarily for non-radioactive testing of plant processes and for 

the development of new processes. As with the High-Bay Lab, some projects were carried out with

un-irradiated uranium. 

The north and south halves of the Low-Bay Laboratory were administratively divided into ten 

experimental spaces, each 8-ft. wide. In these spaces were placed experiment modules consisting of a 

structural-steel framework that housed the experimental equipment (CPP Safety Review, 1974). 

5.1.2.11 CPP-640, Hot Pilot Plant/Headend Process Plant

CPP-640, originally known as the Hot Pilot Plant, was initially designed to test new equipment and 

chemical flowsheets in support of the fuel-processing operations in CPP-601. Built in 1961, the 

building was constructed as an empty shell with five shielded test cells, waste collection tanks in two 

vaults at the lowest level of the building, and an open crane loft with space for chemical makeup 

equipment and removal of cell roof hatches. A major modification in the late 1970s added the 

shielded Mechanical Handling Cave in the process makeup area of the structure for processing 

graphite fuels. 

The facility was designed to allow for maximum flexibility of future test assemblies. Cells 3, 4, and 5 

were equipped with two removable shielding walls to allow for possible large tests. Numerous pipe 

slots through the cell walls (shielded with lead bricks when not in use) and cast-in-place pipe 

penetrations provided access for instrumentation, control piping, and wiring. A heavy-duty cart 

capable of transporting heavy shielded fuel casks served Cells 1 and 2. Fuel charging ports were 

available at the top of the other three cells. Until early in the 1970s, a variety of experiments were 

conducted in the facilities of CPP-640; two of these processes, involving electrolytic dissolution of 

stainless steel-clad fuels and combustion of those clad in graphite, were particularly successful and 

soon took over (known as the “headend” process). After 1973, the structure began to be known as the 

Headend Processing Plant, dedicated first to electrolytic dissolution, and then sharing space with the 

graphite combustion process. Eventually (after the time period under evaluation in this report), Cell 5 

of CPP-640 would be devoted exclusively to the electrolytic dissolution process.
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5.1.2.12 CPP-709, Eastside Service Waste (ESSW) Monitoring Station

The ESSW Monitoring Station was constructed (1951-52) about 75 feet east of CPP-601 to house the 

ESSW monitoring equipment. Waste water from floor drains, steam condensate lines, equipment 

cooling jackets, and the process equipment waste evaporator was routed through the basin of this 

building and pumped to an injection well (CPP Plan, 1997). 

5.1.2.13 CPP-734, Westside Service Waste (WSSW) Monitoring Station

The WSSW Monitoring Station was constructed (1950-60) about 150 feet southwest of CPP-601 to 

house the WSSW monitoring equipment. The normally-uncontaminated wastewater from steam 

condensate and equipment-cooling lines was routed to the basin of this building for monitoring and 

then flowed by gravity to an injection well. Upon detection of radiological contamination in the waste 

flow through CPP-734, an alarm signaled CPP plant operators to immediately stop or divert the 

contaminated waste water flow (CPP Plan, 1997).

5.1.2.14 High Level Liquid Waste Tank Farm

The HLLW Tank Farm has been in service (receiving liquid waste until it can be solidified) since 

1954. The Tank Farm consists of eleven 1,135,624-liter (300,000-gallon) tanks, four 113,562-liter 

(30,000-gallon) tanks, four 68,137-liter (18,000-gallon) tanks, and associated equipment for waste 

transfer, monitoring, and control. One of the 1,135,624-liter tanks is designated as a spare and is 

always maintained empty in the event of an emergency. The 1,135,624-liter tanks are constructed of 

stainless steel and housed in underground concrete vaults. The concrete vaults sit on bedrock about 

13.7 m (45 ft) below grade with the top of the vaults about 3.0 m (10 ft) below grade. This distance 

includes approximately 15.5 cm (0.6 ft) of soil over a synthetic membrane. The 113,562-liter tanks are 

constructed of stainless steel, set on concrete pads, and covered with about 3.0 m (10 ft) of soil. The 

68,220-liter tanks are also constructed of stainless steel and are housed inside stainless-steel-lined 

concrete vaults located in the bottom of the Waste Treatment Building, CPP-604. 

The majority of liquid waste stored in the Tank Farm was generated during the first-, second-, and

third-cycle extraction processes. These wastes include high-level wastes that are composed of firstand second-cycle raffinates and intermediate-level wastes that are composed of third-cycle raffinates 

blended with concentrated bottoms from the Process Equipment Waste (PEW) Evaporator. 

Additional wastes stored in the Tank Farm include fluoride- and cadmium-bearing wastes from the 

fluorinel process, decontamination wastes containing fluoride from waste calcining, process salvage 

streams from J-Cell after extraction, and occasional transfers from WL-l04 and WL-105, which are 

part of west side service waste disposal system.

Liquid wastes are transferred to the Tank Farm from various CPP plant areas through underground 

stainless steel lines. These buried lines and associated equipment constitute both a system for the 

transfer of high-level, liquid radioactive waste and one for the transfer of less-radioactive 

intermediate-level, liquid waste. All the transfer lines are piped and valved separately throughout the 

Tank Farm area so the integrity of the various waste types is maintained. 
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Most of the nonradioactive, liquid discharge originates from the CPP utilities areas in CPP-606, 

CPP-644, and the coal fired steam-generating facility. Cooling water for the secondary heat transfer 

system for the CPP liquid waste Tank Farm is routed directly to service waste without monitoring for 

radioactivity because the primary systems are monitored. Potentially radioactive, liquid effluents at 

the CPP are associated with heating- or cooling-water streams from various processes throughout the 

plant. 

5.1.2.15 Calcined Solids Storage Facilities (CSSF)

Initially, the radioactive liquid waste from the HLLW Tank Farm was sent as feed to the Waste 

Calcining Facility and after the WCF was shut down, to the New Waste Calcining Facility (NWCF). 

The product of the calcining facilities takes the form of fission-product salts that are then transported 

by air to the CSSFs for storage. 

There are seven CSSFs. CSSFs 1–6 provide containment for the radioactive calcined solids. The 

CSSFs are separate (separated by distance), self-contained facilities and there are no interconnections 

between them, except for the storage bin vent off-gas systems. CSSF 7 was completed, but never 

placed into service since CSSF 6 was never filled to capacity. CSSFs 1–6 were sequentially built 

between 1959 and 1984 and placed in service as they were completed. The storage bins of CSSFs 1–5 

are effectively filled to their usable operating capacity. The CSSF 6 storage bins are filled to 

approximately 50% of capacity. CSSFs 1–3 store calcine produced by the WCF. CSSFs 4 and 5 are 

filled with calcined material produced by the NWCF. 

5.1.2.16 Auxiliary CPP Facilities (Non-Radiological)

To support operations, there were several non-radiological support buildings in the CPP complex. 

North of the Laboratory building was the Service Building (CPP-606), which supplied the heat, water, 

air, and emergency power to both CPP-601 and CPP-602. To maintain security, the Main Guard 

House (CPP-609) was erected at the perimeter fence directly west of the Laboratory building. North 

of the Spent Fuel Storage building stands the North Guard House (CPP-610). Completing the list of 

buildings in this area are a number of buildings for storage of non-radioactive parts and materials 

(CPP-607, -608, -615, -616, -617, and -710), electrical transformer buildings (CPP-704, -705, -706, -

707, and -718), pump houses (CPP-611 servicing Well Number 1, and CPP-612 servicing Well 

Number 2). Also present at various times were several temporary structures of frame construction 

with tar paper coverings for service use (CPP-618, -622, -623, -624, -625, and -626). 

5.1.2.17 Overview: Fuel Reprocessing Operations

The CPP fuel-reprocessing operation basically followed five main steps: 

1. The first step involved the transfer of actual spent fuel elements to the plant. These highlyradioactive elements were brought to the CPP in lead shipping casks transported via truck or 

railcar. They were received by a large crane in the fuel storage building, CPP-603, where they 

were stored for days, months, or even years until enough fuel of a particular type had accumulated 

to make a processing run economical. No fresh fuel was accepted at CPP-603; all had a minimum 

of ninety days cooling. Prior to arriving at the CPP, even ninety-day-old fuel had to be stored in 
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CPP-603 for at least thirty additional days in order to meet the minimum 120-days cooling time 

before being submitted for reprocessing. 

2. Once the fuel assemblies stored in CPP-603 had cooled sufficiently, and a large-enough quantity 

of a specific fuel type had accumulated, specific amounts could be placed into a shielded cask and 

loaded onto a specialized Gerlinger straddle truck for transfer to the fuel processing area of the 

plant. Typically this transfer involved a trip of approximately one-third of a mile down Maple 

Street from the Fuel Storage Building, CPP-603, to the Main Processing Building, CPP-601. 

Fuels clad in graphite and stainless steel, and processed after 1961, were also transported down 

Maple Street, but their final destination was the Hot Pilot Plant/Headend Process Plant, CPP-640, 

constructed adjacent to the main processing building. After 1986, some fuels alloyed with 

zirconium were stored and initially processed in a new facility, the Fluorinel Dissolution Facility 

(CPP-666). Chargers, the specialized transfer casks used at CPP, were capable of discharging fuel 

directly to a dissolver, to a shielded chute leading into a dissolver, or to a remote cave or cell, from 

which the dissolver could be loaded remotely with manipulators.

3. Once in the dissolver vessel, the fuel was combined with various acid-based chemical reagents 

that would ultimately transform it into liquid. This aqueous mixture was transferred by steam-jet 

suction to a series of other vessels for separation of the fuel constituents. 

4. Separation took at least two cycles, during which the uranium was extracted from the acids, 

structural alloys, and fission products, through the use of organic solvents. 

5. The uranium solution (uranyl nitrate) was separated from the solvents, which were often purified 

and recycled for new processing runs. 

As in any processing plant, sample analyses form the basis for control of the CPP process; thus,

samples were taken of each of the process streams at varying intervals (CPP Contract, 1955). The 

radioactivity of the various sampled solutions complicated the sampling procedure. All process 

solutions samples had to be removed from the process without exposing personnel. Fission products 

in the sample solution made contamination of the sampling equipment and sampling area a problem 

during plant operation. Designers of the remote sampler and the sample transfer shield considered 

these problems and tried to minimize them as much as possible.

The remote sampler draws the sample from the process stream inside of the cell and discharges it into 

a small glass bottle. A sample line and a sample return line penetrate the cell wall connecting the 

sampler with the vessel to be sampled. In some places, a small vessel called a sample pot is used 

specifically for sampling. In many other cases, the sample is taken directly from the vessel. An air jet 

provides the force required to move the sample stream through the sampling lines. 

The sample and jet are located inside a four-inch-thick lead and steel shield that runs the length of the 

sampling corridor. A four-inch-thick lead glass window penetrates the shield at each sample station 

so that the operator can observe the sampling operation. Architecturally, a bulge in the sampling 

corridor dedicated to sampling was called a “sample blister.” Each sample blister was equipped with 

a remote handling device, a light, and a shielded area (“sample pig garage”) where the sample could 

be transferred into a transport shield (“sample pig”) (Stroschein, 1967). A sample blister might 
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contain multiple sample stations for redundancy in the case of equipment problems and/or 

contamination issues. 

Denitration Process

For many years, the highly-concentrated uranium solution was bottled and then shipped as liquid 

uranyl nitrate solution to Oak Ridge to be further refined and manufactured into reactor fuel. To 

maintain a safe geometry, the stainless steel product bottles were contained in bird-cage-type 

containers. 

Because CPP liquid-shipping containers did not meet all requirements of the safety regulations for 

off-site shipments, as set forth in AEC Manual Chapter 0529, alternative methods of packaging and 

shipping were investigated. Results of the study indicated that conversion of the liquid product to a 

solid UO3 product by fluidized bed denitration, and shipment of the UO3 in standard Y-12 Model-FD 

shipping containers would satisfy the requirements of the AEC Manual. 

All product shipped after 1969 was converted to solid oxide using the following process before 

leaving CPP: 

The uranyl nitrate solution contained in one bank of storage cylinders in Z cell - about 140 liters at 

approximately 350 grams uranium per liter - was processed as a batch. The solution was piped at a 

rate of about 10 L/h to the process room, where an air-atomizing nozzle sprayed the solution into the 

heated fluidized bed. The product denitration process was located in a room built next to the original 

product room in part of an area that in previous years was occupied by a storeroom. The processing 

room was 13 by 18 by 14 feet tall, and had an operating pit, 8 by 6 by 3 feet deep to supply the 

necessary headroom. The room housed all of the necessary control panels and most of the major 

pieces of equipment for the process. The feed line came from Z-cell storage tanks, and the off-gas 

line returned to Z-cell VOG piping after being filtered. The filtering section of the denitrator vessel 

contained three sintered metal filters for removal of oxide dust from the off gases. Each filter had a 

double-filtering wall in a concentric cylindrical shape, and a mean pore diameter of 15 microns. The 

sole purpose of the operating pit was to provide operating room below a glove box used in product 

packaging. 

Sampling and packaging of the granular product was performed in a glove box where each 

60-kilogram batch of oxide produced from denitrating a batch of feed from one Z-cell bank is divided 

among five or six plastic bags. Only one bag of product was normally in the glove box at a time. The 

glove box used for packaging product and taking samples was constructed of stainless steel sheet and 

Plexiglas, and was approximately 26 inches high, 48 inches wide, and 28 inches deep. It contained 

three glove ports, a sphincter, a bag-out port, a door, and a twin-shell blender. It was connected to the 

hood-exhaust system and kept under a slight vacuum. The glove box off-gas line had a highefficiency filter to keep any oxide dust from entering the hood exhaust system (Bjorklund, 1970).
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Waste Storage

Every kilogram of purified uranium produced at the CPP created on average over one hundred gallons 

of acidic liquid radioactive waste that had to be safely stored. These raffinates were maintained in an 

acidic solution and stored in a series of stainless steel tanks constructed at the Tank Farm. Initially, 

only two 300,000 gallon tanks were located there, but eventually, nine more of these large vessels 

were constructed along with the four smaller 30,000 gallon tanks. The large tanks stood 21 feet tall at 

the eaves, 50 feet in diameter, were made of stainless steel, and were enclosed in concrete-lined 

vaults. More than half of them were equipped with cooling coils necessary to safely store intensely 

radioactive solutions.

5.1.2.18 Overview: Isotope Recovery

While the radioactive isotopes that contaminate spent reactor fuel tend to be a radiological nuisance, 

they can be a source of valuable radioactive by-products. At CPP, several were isolated during the 

fuel-reprocessing operation and shipped to other laboratories for use in research and weapons 

production. 

RaLa Process

Lanthanum-140 is produced when barium-140 undergoes radioactive decay. Recovery of La-140 was 

a complicated process completed in a specialized processing cell, L-Cell, in CPP-601. Beginning in 

July 1954, a shielded viewing window, remote-handling manipulators, a periscope, and intercom 

system were installed in the vacant L-Cell that would allow an operator to see and hear the work going 

on behind six feet of concrete shielding. A wide variety of equipment also began to appear inside the 

cell. Equipment added to the cell included an elevator, two large centrifuges, a dissolver and its 

charging chute, an off-gas scrubber, and process tanks of many sizes (Stacy, 2006). 

Construction was essentially complete by November 1, 1955, allowing Phillips Petroleum Company 

to initiate start-up of the extraction program. Calibration of instruments and equipment continued 

until April 1956. Tests using un-irradiated material were conducted between April 25 and November 

23, 1956. Test batches of irradiated material were processed between November 24, 1956 and June 

30, 1957, at which time the process development was considered complete (Legler, 1957) 

The complexity of this project was due in large part to the extremely high amount of radioactivity 

present in the fuel elements. The short-lived barium isotopes were present in irradiated fuel removed 

from the Materials Test Reactor (MTR), but they decayed rapidly after being removed from the 

reactor core. After about forty days, these isotopes were gone. To isolate these short-lived products, 

fuel elements were processed immediately after being removed from the MTR. 

These intensely-radioactive fuel elements were placed in a hot sodium hydroxide solution that 

dissolved the aluminum cladding, leaving the barium and other fission products as solids. This 

combination, or slurry of solids and liquids (somewhat akin to a mixture of mud and water) was sent 

to one of two centrifuges where the whirling motion drove the barium with other solids to the wall of 

the bowl and the clear liquid, with its dissolved metal, to the center where it could be skimmed off. 

Successive dissolution with other reagents, centrifuging and skimming left barium nitrate only on the 

wall of the centrifuge bowl. It was then dissolved in water so that it could be moved to a product cup 
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by skimming. Here it was evaporated to dryness. The cup was placed in a shielded shipping 

container with lead thick enough to shield against the tens of thousands of curies of radioactive 

material from the tiny source (a gram or so). This product was shipped to other AEC labs for 

specialized uses requiring high radioactivity and rapid decay. The RaLa process campaign ran from 

1957 to 1963, during which over a million curies of Ba-140 were recovered. 

Neptunium and Plutonium

For most of CPP’s processing history, the small amounts of neptunium and plutonium present in plant 

dissolver products were sent, with other fission products as waste, to the Tank Farm. From 1965 to 

1972, the first-cycle extraction raffinate streams that contained these isotopes were resubmitted to the 

first-cycle extraction equipment along with chromic and nitric acids, which reoxidized the neptunium 

so that extraction could eventually take place. For seven years, these neptunium/plutonium-bearing 

solutions were stored in tanks in CPP-601 (N-Cell), awaiting a final processing run. In 1972, an 

increased demand for neptunium, combined with an immediate need for more storage space in N-Cell, 

drove a three-week campaign to separate the neptunium and plutonium from the remaining fission 

products through a further extraction cycle. The product solution that resulted, containing 5,412 

grams of neptunium and 544 grams of plutonium, was pumped from the CPP-601 processing 

equipment to a temporary packaging facility in the CPP-627 Multi-Curie Cell. There it was bottled, 

packaged, and shipped in specialized containers to the Savannah River Site for further purification and 

ultimate conversion to Pu-238. 

Rare Gases 

Krypton and xenon are both produced in significant quantities during uranium fission in a nuclear 

reactor. Because of their inert, gaseous nature, both are of possible value to scientific research. To 

recover and purify these gases, a Rare Gas Plant was installed at CPP in 1958. The plant was 

designed to recover several thousand curies per day of krypton, and a significant volume of xenon, 

from off-gas evolving from the acid dissolution of spent fuel elements. The original process used 

charcoal absorbers (operating at liquid nitrogen temperatures) for removing radioactive krypton from 

the dissolver off-gas. When the absorbent was saturated, it was heated to drive off the rare gases 

which were then collected in a cold trap. However, the product from this early process was impure, 

and the cooling requirements of the process required more nitrogen than the existing liquid nitrogen 

plant could produce on a continuous basis. After it became desirable to operate the fuel dissolvers on 

a continuous basis, a cryogenic distillation process was installed to reduce the liquid nitrogen 

requirements and to produce pure products (Offutt, 1969). 

5.1.2.19 Overview: Waste Calcination

The WCF began operations on May 18, 1961 with two years of "cold" operations using simulated 

waste. Cold start-up consisted of 10 runs: the first run used water as feed to check out equipment and 

operational problems; the other runs used simulated Al(NO3)3-nitric acid waste to check out 

equipment, operational, and calcination flowsheet problems (Newby, 2000). Such runs illuminated 

numerous deficiencies in the equipment or the process, all of which were overcome or accounted for 

in various modifications and adjustments. In addition, the cold runs provided analysts an opportunity 

to determine what safety precautions were needed to operate the facility. The two years of 
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experimental runs prepared the plant for its first hot run, which began on December 23, 1963 (INEEL, 

1998).

Originally conceived by scientists at Argonne National Laboratory, fluidized-bed calcination was 

developed to convert radioactive waste solutions to solids (INEEL, 1998). In the calcination process, 

liquid radioactive wastes are evaporated and the dissolved metals and fission products are converted to 

salts and oxides. The key step in the calcination process is evaporation and solidification of the liquid 

radioactive wastes in a heated bed of fluidized particles. Feed is atomized by air and sprayed into a 

heated bed of spherical particles. Dissolved metals and fission products oxidize to solids as they coat 

the surfaces of bed particles. Combustion and solid products build up in layers on the bed particles 

while the gaseous products are swept from the vessel with the fluidizing gases. Most solids remaining 

in the gas after leaving the primary cyclone are removed in the gas-scrubbing system. 

Before being released to the atmosphere, this stream is decontaminated in a series of devices including 

a cyclone, spray quench tower, scrubber, associated separators, and silica gel absorbers. After 

filtering, the decontaminated off-gas is transferred to the 76.2-m (250-ft) high CPP stack. The spray 

quench tower removes particulates from the off-gas; additional particulate is removed through the 

nitric-acid scrubbing system. Liquid drops from the scrubber are removed to the de-entrainment 

separator and mist eliminator and then drained to the scrub tank and recycled as calciner feed by 

transferring it to the waste holding tanks.

Conversion of HLLW to a solid reduces the volume of waste by a ratio of about 7-to-1 and provides a 

less-corrosive waste form. The calcined solids are in the form of particles 0.2 to 0.6 mm in diameter 

and made up of aluminum and zirconium oxides. Other metallic oxides and fission product oxides are 

withdrawn continuously from the bottom of the calciner and are air-conveyed (pneumatically) to 

underground storage bins. The calciner product removed from the bed and the fines removed from the 

off-gas in the cyclone are combined and pneumatically-transferred to the Calcined Solids Storage 

Facilities (CSSF).

Nine calcination campaigns were completed at the WCF, beginning in December 1963 and ending in 

1981 when the WCF was replaced by the NWCF. After each campaign, the crews set about 

decontaminating the cells so that equipment could be inspected and serviced. Automated sprays and 

steam jets flushed out the cells first, then the technicians - suited-up and masked - raised the hatches 

and descended into the cells to finish the job, replace worn parts, and make repairs. As the facility 

aged, post-campaign reports grew less and less enthusiastic about how easy or safe the hands-on 

maintenance was getting to be. Ruthenium, for example, began to plate out on the interior of the 

off-gas process vessels. It resisted the internal sprays and exposed workers to hazard as they tried to 

remove it manually. 

Up until 1973, the same crew operated the Fuel Processing Complex and the WCF. Each facility 

alternated being “on” and “off,” which gave adequate time for proper maintenance of the standby 

facility. By 1973, demand was so great that the two facilities had to operate simultaneously. Design 

studies also began that year for a new, bigger waste calcining facility. 
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5.1.3 Test Area North (TAN)

ATTRIBUTION: Section 5.1.3 was completed by Brian Gleckler, Dade Moeller, Inc. All conclusions 

drawn from the data regarding Test Area North were peer-reviewed by the individuals listed on the 

cover page. The rationale for all conclusions in subsequent sections of this document devoted to Test 

Area North are explained in the associated text.

As indicated earlier in Figure 5-2, Test Area North is located in a remote northern portion of the INL 

Site (approximately 30 miles northeast of the CFA). In 1951, General Electric (GE) was awarded 

United States Air Force and Atomic Energy Commission contracts to perform the direct cycle work 

for the Aircraft Nuclear Propulsion (ANP) Program (Stacy, 1995; Thornton, Feb1962). Ground was 

broken to construct TAN in February 1953, and by Christmas 1955 serious operations were underway 

at TAN (Stacy, 1995). The TAN facilities were collectively referred to as the Idaho Test Station 

under the GE contracts. On March 28, 1961, the ANP Program was terminated (Thornton, Feb1962). 

The ANP Program activities at the TAN were replaced by Safety Test Engineering Program (STEP). 

STEP received AEC approval on December 28, 1961. On February 6, 1962, the program was 

assigned to the prime contractor for the INL Site, Philips Petroleum Company. The overall objectives 

of STEP were: to conduct full scale tests on reactor systems to determine what actually happens to 

fission products released during a reactor accident; to demonstrate reactor safety; to develop realistic 

analytical methods useful for future reactor designs; and to use basic information already developed in 

the Nuclear Safety Program (INL, 1963). 

The major facilities at TAN included:

• Technical Support Facilities (TSF)

• Initial Engine Test (IET) Facility (renamed the Initial Engineering Test Facility during the STEP 

era)

• Flight Engine Test (FET) Facility (renamed the Loss-Of-Fluid Test (LOFT) Facility during the 

STEP era)

• Low Power Test Facility (LPTF) Area

The physical locations of each of these facilities are depicted in Figure 5-27. The TAN contained 

approximately 57 buildings during the period for the SEC petition (INL, 2009). Table 5-4 lists the 

notable radiological structures at the TAN.
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Figure 5-27: Map of Test Area North (TAN)
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Table 5-4: Notable Radiological Structures at the TAN

Facility Name Building No. First Used Last Used Facility Type

TSF Area

A&M Building TAN-607 1955 beyond EP Assembly, maintenance, 

and laboratory facility

Hot Cell Annex (HCA) TAN-633 1955 beyond EP Hot cells and laboratory

Hot Liquid Waste Treatment Plant TAN-616 1955 beyond EP Waste treatment facility

Actuator Building and later TAN 

Fuel Handling Facility TAN-615 1956 beyond EP

Actuator testing facility and 

later a reactor fuel handling 

facility

Dolly Storage Building TAN-647 1960 beyond EP Radioactive materials 

storage facility

Parts Storage Building TAN-648 1960 beyond EP Radioactive materials 

storage facility

Health Physics Building TAN-606 ND beyond EP Laboratory and radioactive 

source storage

AEC Health Physics Building TAN-671 ND ND Laboratory and radioactive 

source storage

TSF Gatehouse TAN-601 1956 beyond EP Dosimeter exchange 

location

IET Facility

Control and Equipment Building TAN-620 1955 1966 Reactor control building

Test Cell Building TAN-624 1955 ND Movable weather protection 

building for experiments

Coupling Stationb NA 1955 ND

Coupling station for 

exhaust from nuclear 

powered jet engines

Change Room TAN-656 1960 ND Room for donning and 

doffing protective clothing

IET Guardhouse TAN-621 1955 beyond EP Dosimeter exchange 

location

IET Exhaust Stack TAN-712 1955 ND 150 ft exhaust stack for 

reactor experiments

FET/LOFT Facility

Test Building (a.k.a. Hangar 

Building) TAN-629 1959 beyond EP Storage facility

LOFT Reactor TAN-650 beyond EP ND Reactor

Control & Equipment Building TAN-630 beyond EP ND Reactor control building

LPTF Area

LPTF Control & Equipment Building TAN-641 1958 early 1970s Reactor control building

LPTF Assembly & Test Building TAN-640 1958 early 1970s Reactor test facility

STPF Control & Equipment Building 

and later EBOR Facility TAN-645 1960 1966 Reactor control building

STPF Pool Facility Building and later 

EBOR Facility TAN-646 1961 1966 Reactor test facility

ND = Date of use not determined

NA = Not available

beyond EP = beyond evaluation period
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5.1.3.1 Technical Support Facilities (TSF)

The TSF was originally built to provide technical and administrative support for the ANP Program. 

The TSF was later adapted to support other programs and projects, such as STEP. The TSF consisted 

of two areas: the eastern portion was the Assembly and Maintenance (A&M) Area; the western 

portion was the Administrative Area (INEEL, 2005). The TSF was comprised of a wide variety of 

structures that were used for radiological and non-radiological purposes, such as: warehouses, 

administrative buildings, guardhouses, change houses, buildings for utilities, waste handling and 

disposal facilities, maintenance shops, assembly areas, hot cells, steam plant, jet engine test building, 

and a large reactor assembly and maintenance building (AEC, 1954-1986; AEC, 1954-1960; 

Stroschein, 1967; INEEL, 2005). All of the radiological areas were within the A&M Area at the TSF. 

The significant radiological areas in the A&M Area included the A&M Building, Hot Cell Annex, Hot 

Liquid Waste Treatment Plant, Actuator Building, and the Radioactive Parts Security Storage Area. A 

map of the TSF is provided in Figure 5-28. 
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Figure 5-28: Map of the Technical Support Facilities (TSF) 
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A&M Building (TAN-607)

The A&M Building was approximately a 200 foot by 500 foot, two-story, reinforced concrete 

building that was constructed in 1954 (Picker, 2001; INEEL, 2005; LOC, 1957-2014). The major 

parts of the building included the Hot Shop, Warm Shop, Special Equipment Services (SES) Room, 

Radioactive Materials Laboratory, Storage Pool, Decontamination Room, High Bay Assembly Shop, 

Metallurgical Laboratory, radiochemistry laboratories, craft shops, and administrative areas (Freeman, 

1964; INEEL, 2005; EG&G, 1977). Figure 5-29 presents the floor plan for the A&M Building.

The A&M Building’s Hot Shop was the world's largest hot shop, measuring 51-feet-wide, 165-feetlong, and 55-feet-high. Its outer concrete walls were seven-foot-thick with six-foot-thick viewing 

windows. The Hot Shop was the hub of the A&M Building and was designed to provide a remote 

maintenance and disassembly area where several manipulators operate together in conjunction with 

turntables, fixtures, and standard and special tools on the largest and smallest radioactive industrial 

components. The Hot Shop was also equipped with a 100-ton bridge crane and an Overhead 

Manipulator (O-Man). Heavy equipment could be brought to this area on rails leading from the 

outdoor turntable (Freeman, 1964). 

The Warm Shop was located adjacent to the Hot Shop, and was a high-ceiling area where test 

assemblies with minor activity or contamination could be stored and worked on. The Warm Shop 

facilitated rapid handling of slightly-radioactive (or warm) components. Personnel, wearing 

protective clothing and working under controlled environmental conditions, were able to perform the 

same functions of handling, observation, and analysis as performed by the remote facilities in the Hot 

Shop. As with the Hot Shop, heavy equipment could be brought to this area on rails leading from the 

outdoor turntable (Freeman, 1964). 

Located at the back of the Hot Shop was the SES Room, which was a continuation of the upper level 

of the Hot Shop. The SES Room could be closed off from the Hot Shop by means of sliding shielding 

doors. With relative ease, equipment could be decontaminated and repaired in the SES room while 

other operations were being carried out in the Hot Shop. A control station with a mineral-oil-filled, 

shielded viewing window serves to control the equipment being serviced in the SES room (Freeman, 

1964). 

The Radioactive Materials Laboratory (RML), which adjoins the east end of the Hot Shop, is a 

remote-handling area 10-feet-wide by 35-feet-long used for study, observation, and analysis of fuel 

elements and smaller radioactive objects. Access from the Hot Shop was gained through the plug 

door. Delicate, accurate handling and analyzing equipment is employed in the RML. The area is 

serviced by master-slave manipulators at each viewing window and by two bridge-mounted 

manipulators that cover the entire cubicle. Special equipment in the RML included: a remotely 

operated four-kilogram balance having an accuracy of ± 10 milligrams; two Kollmorgen periscopes 

equipped with cameras; a Bausch and Lomb remote microscope; a GEL periscope; an Elox electron 

drill; and miscellaneous remotely operated tools, jigs, and fixtures for analyzing and handling the 

various specialized components (Freeman, 1964). 
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Figure 5-29: Floor Plan of the A&M Building (TAN-607)
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Also adjoining the Hot Shop is a 24-foot-deep storage pool. It is connected to the Hot Shop by an 

extension that passes under a shielding wall. An underwater dolly and track system provides 

transportation to the main pool (Freeman, 1964). 

The Decontamination Room was used for decontaminating parts and equipment, and was located at 

the south end of the A&M Building adjacent to the High Bay Assembly Shop. Its walls were made of 

poured, reinforced (ordinary) concrete that were about 1.5-feet-thick. The room was about 39-feethigh and about 32-feet-wide, and was equipped with a 10-ton bridge crane. Acid and neutralizer pits 

contained fluids used in cleaning operations. Radioactively-contaminated liquid wastes entered an 

underground drain that went to the Hot Liquid Waste Treatment Plant (TAN-616) (INEEL, 2005). 

The High Bay Assembly Shop was maintained in a radiologically-clean condition so that normal 

unobstructed assembly work could be conducted. This shop was serviced by the four-rail track 

system and two separate overhead bridge cranes. Initial assembly of major test components took 

place in the High Bay Cold Assembly Shop, which was also equipped to do large and small 

machining, welding, and special inspections (Freeman, 1964). 

The A&M facility was also equipped with a variety of specialized laboratories. The Chemical Lab 

(second floor) was equipped to analyze samples, certify materials, and study such problems as 

turbidity, which occasionally appeared in viewing glass fluid. It had a balance room and an 

emergency shower. The Metallurgical Lab (next to the Chemical Lab, second floor) had a mobile 

X-ray unit for inspecting welds (first floor), a plating room, a darkroom for developing X-ray film, a 

tensile testing machine, and radiograph-making equipment for producing weld reports. The 

Photographic Lab (first floor) handled standard black/white film and motion pictures. A Radiographic 

Materials Lab included an Elox block cutter with a remotely-operated balance, a periscope/camera 

outfit for making remote images, sample cutters, and mounting equipment. This lab was ventilated 

for negative air pressure and air-filtered with disposable filters (INEEL, 2005). 

Hot Cell Annex (TAN-633)

The Hot Cell Annex (HCA) adjoins the A&M Building’s storage pool opposite of the Hot Shop, and 

is used for the preparation and analysis of radioactive metallurgical specimens (Freeman, 1964). 

These specimens are removed from test assemblies in the A&M Building’s Hot Shop or are brought 

from other parts of the INL Site in casks (Freeman, 1964). The specimens are moved by remote 

control in a straight line from the pool building, down the corridor, and into the HCA. They traveled 

on a 5-ton monorail as far as the designated hot cell. After the hot cell analysis had been completed, 

the item was returned to its transport cask and sent back into the monorail corridor. Here the cask 

could return toward the pool for storage or out a west-side exit to a waiting transport truck. 

Alternatively, the cask could move northward and out the north door of the building. There a concrete 

pad, sheltered by a gabled canopy with a corrugated tin roof, allowed for a transport truck to pick up 

the cask (INEEL, 2005). 

The HCA consists of four shielded hot cells with interconnecting sliding drawers for the transfer of 

samples (Freeman, 1964). The hot cells have mineral oil-filled lead glass windows. Each hot cell was 

equipped with a unique selection of work tables, manipulators, and other equipment. All hot cells had 

air exhausts, floor drains, and a small "lost tool opening" on the east side (INEEL, 2005). 
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Work spaces within the HCA included a change room, a set-up area, and tool decontamination area. 

After the ANP program had been canceled, the room in the northwest corner was converted to a 

photographic darkroom (INEEL, 2005). The mechanical equipment room (Room 10) was in the 

southwest corner. Figure 5-30 provides the floor plan for the HCA. 

Hot Liquid Waste Treatment Plant (TAN-616)

As indicated in Figure 5-28, the TAN-616 was near the north end of the A&M Building. The strategy 

for dealing with radionuclide-contaminated liquids that accumulated due to ANP Program and STEP 

operations, decontamination activities, accidents, and other processes was to isolate, treat, and 

concentrate them in a storage tank. To that end, a complex system of waste drains, pumps, holding 

tanks, and piping connected the A&M Building labs, hot shops, hot cells, and decontamination shop to 

the TAN-616 plant. Waste drains came from the IET Facility, the Actuator Building (TAN-615), and 

the Water Filtration Building (TAN-649) (INEEL, 2005). 

The system involved various treatments for the waste (acid and caustic) and an evaporator to drive off 

water. Initially, the system had three holding tanks (V-1, V-2, and V-3), but another (V-9) was added 

later. The first three tanks were buried underground on the west side of the plant building. These held 

wastes being delivered for treatment and concentration (INEEL, 2005) 

The large evaporator vessel was the central feature of the concentration process. Beginning in 1958, 

sludge from the bottom of the vessel was periodically removed to tanks located southwest of the 

A&M Building in a field on the south side of Snake Avenue. After years of operation, the vessel 

cracked from stress corrosion. The contents leaked onto the floor of the pit. The vessel was removed 

and repaired, but eventually cracked once more. Again, leaks ended up on the floor of the pit. The 

facility was shut down in 1970, but waste in the process system at the time remained where it was. A 

new system, called the PM-2A, was installed, but operating difficulties shut it down in 1975. Again, 

untreated waste remained in process lines (INEEL, 2005) 

Actuator Building (TAN-615)

As indicated in Figure 5-28, the Actuator Building was just outside the north end of the Hot Liquid 

Waste Treatment Plant. GE designed the Actuator Building in November 1956 and built it soon after 

(INEEL, 2005). For the ANP Program, the Actuator Building served as the location for testing 

prototype actuator mechanisms, the motors, instruments, electromagnets, and other devices that 

moved control rods in and out of the HTRE reactors (INEEL, 2005). The TAN-615 was later 

renamed the TAN Fuel Handling Facility, but the date when that transition occurred is unknown 

(Waste Management, 1974, PDF p. 373). In March 1963, the TAN-615 was handed over to the 

Phillips Petroleum Company (Phillips), which had the STEP contract. Upon turnover, Phillips 

performed a thorough inspection and found the entire building to be highly contaminated with alpha 

radioactivity (TAN HP Report, Dec1963, PDF pp. 45-49). The alpha radioactivity was later 

determined to be from uranium (Dierks, 1963). 
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Figure 5-30: Floor Plan of Hot Cell Annex (TAN-633)
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Radioactive Parts Security Storage Area (TAN-647 and TAN-648)

Storage buildings (TAN-647 and TAN-648) and the adjacent storage pads are known as the 

Radioactive Parts Service and Storage Area (RPSSA). These buildings are in the northwest portion of 

the TSF area. The area has residual contamination from earlier projects, including the HTRE-2 and 

HTRE-3 reactors. The buildings and contaminated areas are marked with perimeter fencing and 

warning signs. The gates and buildings are locked when access is not required (ORAUT-TKBS-0007-

2). 

Access to the RPSSA was available either by road or by rail, depending on the item being stored 

(INEEL, 2005). Dollies went via rail on a new set of tracks leading from the turntable directly into 

TAN-647, named the Dolly Storage Building. Access by road was from an existing road leading to 

the north end of the A&M Building. Items could be stored inside TAN-648, named the Parts Storage 

Building, or set outdoors. An area southwest of TAN-647 was designated for transport cask liners. 

(INEEL, 2005). 

5.1.3.2 IET Facility

The IET Facility was located approximately one mile north of the TSF. When this facility was built 

for the ANP Program, it was named the Initial Engine Test (IET) Facility, but was later renamed 

Initial Engineering Test (IET) Facility during the STEP era (Ref: PTR-589 – Not Yet In SRDB). 

Figure 5-31 provides an aerial photograph of the IET Facility. Figure 5-32 provides a map of the IET 

Facility. 

The IET Facility was built as a shielded test facility to prove that heat from a nuclear energy source 

could run a turbojet engine for the ANP Program (Stacy, 2000). The IET Facility included shielded 

underground control and equipment rooms, jet engine fuel storage, water-cooling facilities, and an 

effluent exhaust gas system. The facility was unique among other reactor installations in that the 

radiation shielding enclosed the personnel rather than the reactor, which made the facility very 

flexible for adapting to testing and operating power plants of varying size. The power plant could be 

viewed directly from the control room through either of two periscope eye-pieces. Personnel could be 

dropped off or picked up within 100 feet of the power plant by a shielded locomotive. The 

locomotive was equipped with a shielded retractable hatchway in the floor of the cab that would mate 

with a hatch that was between the two center rails. Personnel could then travel to and from the control 

room via an underground tunnel (General Electric, 1953). 

Three U-235-fueled nuclear reactor assemblies, called Heat Transfer Reactor Experiments (HTREs), 

were operated at the IET Facility during the ANP Program era (DOE, 1991; Thornton, Jun1962). The 

three HTRE assemblies were designated HTRE-1, HTRE-2, and HTRE-3 (Thornton, Jun1962). The 

HTRE assemblies were mounted on a four-track railroad dolly (DOE, 1991). Each test for the HTRE 

assemblies was designated with an Initial Engine Test (IET) number, and some of the IETs would 

have multiple runs (DOE, 1991). After an IET run, the shielded locomotive hauled the HTRE 

assemblies back to the TSF Hot Shop in the TAN-607 building, where specialists were ready to 

examine, refuel, maintain, and repair the reactor and engine (INEEL, 2005). The HRTE-1 reactor 

went critical for the first time on November 4, 1955, for tests prior to connection with the turbojet 

engines (Stacy, 1995; Thornton, Feb1962). On December 30, 1955, the HRTE-1 reactor went critical 

for the first time with the turbojet engines connected to it (Stacy, 1995). Tests involving the HTRE-1, 
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HTRE-2, and HTRE-3 reactor assemblies went on until the ANP Program was cancelled in March 

1961 (DOE, 1991). 

On December 28, 1961, the AEC initiated another nuclear safety program that used the IET Facility; 

that program was called the Safety Test Engineering Program (STEP) (INL, 1963). The portion of the 

STEP that used the IET Facility was the reactor test series called Systems for Nuclear Auxiliary 

Power (SNAP) Transients (SNAPTRAN) Program. During the SNAPTRAN Program, the Initial 

Engine Test (IET) Facility was renamed Initial Engineering Test (IET) Facility (INL, 1963; Ref: PTR589 – Not Yet In SRDB). 

The SNAPTRAN Program was designed to provide nuclear safety information applicable to 

SNAP 10A/2 reactor systems (AEC, Apr1967). The SNAP 10A/2 reactors were zirconium-hydrideuranium (ZrHx –U) fueled, sodium-potassium alloy (NaK) cooled, and beryllium-reflected (Bentzen, 

1963). The program consisted of the SNAPTRAN-1, -3, and -2 test series, which were conducted in 

that order. Each of these tests investigated the transient behavior of the SNAP 10A/2 fuel under largetransient, power-excursion conditions (AEC, Apr1967). The SNAPTRAN reactors were mounted on 

a four-rail flatcar so that the reactors could be returned to the TAN Hot Shop in the TAN-607 building 

for inspection and investigation (DOE, 1991). The SNAPTRAN-1 tests investigated the 

nondestructive transient reactor response to large-reactivity additions. This series of tests was 

conducted under conditions approaching, but not resulting in, fuel damage (AEC, Apr1967). The 

SNAPTRAN-3 destructive test investigated the transient reactor behavior and the radiological 

consequences of reactor destruction under conditions simulating accidental water immersion (AEC, 

Apr1967). The SNAPTRAN-3 test was conducted at 11:44 a.m. on April 1, 1964, with extensive 

radiological and meteorological support, surveillance, and controls (DOE, 1991). The SNAPTRAN-2 

portion of the program investigated the dynamic behavior of the zirconium-hydride-uranium fuel 

under the influence of reactivity additions large enough to produce destruction of the core (AEC, 

Apr1967). The SNAPTRAN-2 test was conducted at 9:51 a.m. on January 11, 1966 with extensive 

radiological and meteorological support, surveillance, and controls (DOE, 1991). 
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Figure 5-31: Aerial Photograph of the IET Facility (facing east)
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Figure 5-32: Map of the IET Facility
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5.1.3.3 Flight Engine Test (FET)/Loss-Of-Fluid Test (LOFT) Facility

The FET/LOFT Facility was located approximately 1.15 miles west-northwest of the TSF. A map of 

the FET/LOFT Facility is provided in Figure 5-33. 

The FET Facility was originally constructed to serve as an aircraft hangar facility for a huge nuclear 

powered airplane, as part of the ANP Program (Stroschein, 1967). The primary structure in the FET 

Facility was the aircraft hangar, the TAN-629 building. The TAN-629 building is a large structure 

that measures 240 feet by 320 feet, and has a maximum height of 99 feet. Because a nuclear-powered 

airplane was never built before the cancellation of the ANP Program, the FET Facility was never used 

for its designed purpose. After the ANP Program, the TAN-629 building was used as a “clean” 

storage area, and health physics work in that area normally consisted of surveying materials going in 

and out of the building for possible contamination (Stroschein, 1967). 

For a relatively brief period (approximately 1962–1964), the FET/LOFT Facility was referred to as 

the STEP Area. During that period, little radiological work was performed at this area, and that work 

was likely limited to the storage of materials in TAN-629. 

In 1964, construction began on the LOFT reactor (INEEL, 2005). Sometime after that, the FET 

Facility was renamed the LOFT Facility. The LOFT reactor at the TAN-650 building was a 

centerpiece in the safety testing program for commercial power reactors. The reactor was a scale 

model of a commercial pressurized water reactor (PWR) that was built to explore the effects of Lossof-Coolant Accidents (LOCAs). Thirty-eight nuclear power tests were eventually conducted at the 

LOFT Facility with various accident scenarios, including the accident at Three Mile Island (TMI). 

Among other goals, the program investigated the capability of emergency core cooling systems to 

prevent core damage during a LOCA. Experiments at LOFT simulated small-, medium-, and largebreak LOCAs, sometimes complicated with other events such as loss of off-site power. However, the 

first nuclear test at the LOFT reactor did not take place until December 10, 1978, which is outside the 

period under evaluation for this SEC report (INEEL, 2005). 
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Figure 5-33: Map of the FET/LOFT Facility
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5.1.3.4 Low Power Test Facility (LPTF) Area

The LPTF Area, which included the LPTF and the Shield Test Pool Facility (STPF), was 

approximately 1.25 mi south-southeast of the TSF. A map of the LPTF Area is provided in Figure 

5-34. A photograph of the LPTF Area is provided in Figure 5-35. The first facility at LPTF Area was 

the LPTF in 1958. The Shield Test Pool Facility (STPF) in the west side of the LPTF Area was built 

as part of the ANP Shielding Experimentation Program; in 1963, the pool facility was later modified 

for the Experimental Beryllium Oxide Reactor (EBOR). Note the photo in Figure 5-35 is dated later 

than the map in Figure 5-34; the far left of the photo shows the additions to the STPF that were added 

for EBOR. 

Figure 5-34: Map of the LPTF Area
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Figure 5-35: Photograph of the LPTF Area (looking northwest)

Low Power Test Facility (LPTF)

The LPTF was used to conduct several low-power (less than 100-W) reactor research programs during 

the period 1958–1973. The LPTF contained two shielded cells with three independent control rooms 

and necessary support facilities. The north cell was called the Critical Experiment (CE) cell (Room 

101); the south cell was the Initial Criticality (IC) cell (Room 101). The test cells are made of poured 

concrete construction with a 4-feet-thick wall between them. The walls between the cells and the 

control room are 5-feet thick and 30 feet high. The outside wall of the IC cell is 2-feet thick, and the 

outside wall of the CE cell is 3-feet thick and 30 feet high (Kunze, 1970; ORAUT-TKBS-0007-2). 

The construction of the facility was such that more than one reactor program could be running at the 

LPTF at the same time. Heavy experiment pieces could be moved in or out of each cell through large 

roll-up doors on the east side of the LPTF (ORAUT-TKBS-0007-2). 
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The seven nuclear reactors that operated at the LPTF are listed in Table 5-5 below, all of which were 

uranium-fueled reactors (Stacy, 2000; Becar, 1961; INEEL, 2005).

Semiscale also operated at the LPTF between 1970 and 1971 (INEEL, 2005). Semiscale was a 

non-nuclear apparatus that simulated the heat of a nuclear reactor to study thermal-hydraulic 

phenomena during postulated loss-of-fluid accidents. 

Table 5-5: Nuclear Reactors that Operated at the LPTF

Nuclear Reactor Name Abbreviation

or Acronym First Used Last Used

Cavity Reactor Critical Experiment CRCE 05/17/67 early-1970s

Critical Experiment Tank CET 1958 1961

Fast Spectrum Refractory Metals Reactor 710 Reactor Mar 1962 1968

High Temperature Marine Propulsion Reactor 630-A Reactor 1962 1964

Hot Critical Experiment HOTCE 1958 1961

Spherical Cavity Reactor Critical Experiment SCRCE Nov 1972 1973

Thermal Reactor Idaho Test Station THRITS 1964 1964

Source: Stacy, 2000; INEEL, 2005

Shield Test Pool Facility (STPF)/Experimental Beryllium Oxide Reactor (EBOR)

The STPF was initially built to perform shielding experiments in support of the ANP Program 

(INEEL, 2005). The STPF was designed to do radiation-scattering tests in water utilizing the Susie 

Reactor as its radiation source (INEEL, 2005). The Susie Reactor was a low-power swimming pool 

reactor that had a fuel loading of approximately 4 kg of aluminum-clad U-235 (General Electric, 

1961; INEEL, 2005). STPF had two pools to allow testing to proceed in one while mock-up work 

could be done in the other. 

After the ANP Program was cancelled in 1961, the Susie Reactor was utilized for other projects at the 

INL site (INEEL, 2005). However, by the end of 1962, the Susie Reactor was no longer in use. In 

June 1963, construction began to adapt the STPF for the EBOR, which involved removing the Susie 

Reactor and draining the STPF pools (INEEL, 2005). In December 1966, the EBOR project was 

cancelled before the reactor could ever be loaded with nuclear fuel. 
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5.1.4 Misc. Reactor Areas (ARA, OMRE, SPERT) 

ATTRIBUTION: Section 5.1.4 was completed by Mitch Findley, MJW Corporation. All conclusions 

drawn from the data regarding the Misc. Reactor Areas were peer-reviewed by the individuals listed 

on the cover page. The rationale for all conclusions in subsequent sections of this document devoted 

to the Misc. Reactor Areas are explained in the associated text.

5.1.4.1 Auxiliary Reactor Area (ARA)

The ARA was initially known as the Army Reactor Experiment Area (AREA), then the Army Reactor 

Areas (Redesignation, 1963) before becoming known as the Auxiliary Reactor Area. In February 

1954, the Army was assigned the task of “developing nuclear power plants to supply heat and 

electricity at remote and relatively inaccessible military installations.” In 1955, the Department of 

Defense notified the AEC of the need for a nuclear power plant with the following initial criteria: 

constructed of air-transportable components, and capable of supplying about 200 kilowatts of 

electricity and 400 kilowatts of heat. These plants would be used at military sites to supply power for 

radar equipment, and heat for offices, barracks, and other buildings. A military-oriented program 

began in 1957 at INL to design and test a mobile power reactor that was easily transported and 

operated with minimal maintenance and technical support (Stacy, 2000). ARA was located on 20 

acres with four separate operational areas: ARA-I, ARA-II, ARA-III, and ARA-IV.

Construction on ARA-I began in 1957. ARA-I was built to support the Stationary Low Power Reactor 

(SL-1) which was located in ARA-II. ARA-I was strictly a support facility which included two hot 

cells (Building 626) (see Figure 5-36) and a maintenance/metallurgical shop (Building 627) (AEC, 

1954-1986). ARA-I was initially used in direct support of the Army reactors for examining fuel 

elements and preparing irradiated specimens for evaluation. Beginning in January 1966, ARA-I was 

used for a variety of technical support services for INL’s developmental research programs. 

Examples include the Test Reactor Area neutron cross-section program and separation of some highactivity radionuclides. In 1968, the ARA hot cell was extensively modified for work separating 

Pa-233, a high-specific-activity alpha-emitter (Progress Report, 1968). Modifications included 

upgrades to the ventilation system and addition of isolation boxes for the Pa-233. The facility was 

used intermittently for storage and other miscellaneous reasons until the mid-1980s. 
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Figure 5-36: ARA-I Hot Cell Wall with Viewing Windows and Remote-Handling Manipulators

Construction on ARA-II began in 1957. SL-1 was initially named the Argonne Low Power Reactor 

(ALPR) and was to be the prototype low-power boiling water reactor that was to be a “package” 

power plant at small remote locations, such as the Distance Early Warning (DEW) installations. The 

reactor design was to allow for three years of operation between fuel loadings. ALPR achieved 

criticality at 10:30 pm on August 11, 1958 but was formally dedicated on December 2, 1958, at which 

time it was re-designated SL-1 (see Figure 5-37). After operating intermittently for approximately 

two years, on January 3, 1961 an accident (the first fatal nuclear accident in the U.S.) destroyed the 

SL-1 reactor (SL-1 Report, 1961; SL-1 Records, 2009; SL-1 Chronology, 1961). At that time, ARA-1 

served as the staging area for the emergency response and subsequent decontamination efforts. In 

1961 and 1962, a massive dismantling, disposal, and clean-up of the SL-1 site occurred. The reactor 

building was removed and the support facilities and office buildings were decontaminated. A 

dedicated burial ground was created not far from the reactor facility to minimize external radiation 

exposure as well as contamination-control issues that would have resulted from transporting the 

materials to the INL Burial Ground located approximately 16 miles away (RWMC History, 1985). 
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Figure 5-37: SL-1 in 1960 (Prior to 1961 Accident)

ARA-III was completed in 1959 and was the location of the Army’s Gas Cooled Reactor Experiment 

(GCRE) (GCRE Progress Report, 1959). The GCRE was a water-moderated, nitrogen-cooled lowpower reactor that generated heat but no electricity. It was the first phase in developing a truly mobile 

nuclear power plant and achieved criticality on February 23, 1960. After accomplishing its “proof of 

principle” phase of a mobile nuclear reactor, it was placed in standby mode on April 6, 1961 (Stacy, 

2000). 

ARA-IV housed the Mobile Low Power Reactor No. 1 (ML-1) shown in Figure 5-38. ML-1 followed 

the GCRE’s “proof of principle” by developing an entire plant that could be transported by cargo 

planes or tractor trailers. ML-1 was remotely operated from a control room located approximately 

500 feet from the reactor in order to avoid extensive shielding. The 500-foot distance to the control 

room represented the personnel exclusion zone during reactor operation. The reactor achieved 

criticality on March 30, 1961 but was only operational less than 1000 hours until it was shut down in 

1964 (GCRE Final Report, 1964). 

By late 1965, the Army phased out ML-1 and the entire Army Reactor Project at INL due to questions 

of cost effectiveness, lack of need, and increasing resource needs for the Vietnam War. In mid-1967, 

a small pulsed reactor named FRAN, which was capable of supply bursts of high intensity 

fast-neutron and gamma radiation was transferred to INL from the Nevada Test Site. FRAN was sited 

at the former ML-1 reactor building and was employed for a short time to test the performance of new 

detection equipment being developed for reactor-control purposes (Long, 1969). FRAN achieved 

criticality on August 28, 1968. The reactor was moved to the Lawrence Radiation Laboratory in 

Livermore, California in June 1970. 
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Figure 5-38: ML-1 After Being Moved to ARA-III 
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5.1.4.2 Organic Moderated Reactor Experiment (OMRE) 

A reactor concept conceived and partially-financed by Atomics International, the Organic Moderated 

Reactor Experiment (OMRE) was constructed and operated for several years to demonstrate the 

technical and economic feasibility of using a liquid hydrocarbon, typically a polyphenyl, as both 

coolant and moderator. The primary purpose of the reactor was to study the radiation and thermal 

stability of the organic materials employed, and the associated physical property changes under actual 

reactor operating conditions (OMRE, 1957). The concept of an organically-moderated reactor was 

conceived during the wartime Manhattan Project, but due to the lack of an immediate need for power 

reactors and the general unavailability of enriched uranium, serious consideration did not occur until 

about 1953. 

The AEC had, since approximately 1953, continuously funded (albeit at a modest level) an organic 

reactor development program. Due to the technical uncertainties associated with the use of an organic 

material as a reactor coolant, a very large fraction of the development effort was devoted to the 

determination of coolant properties and stability data. The design and construction of OMRE began 

in mid-1955 because of the recognized need for the test of an organic fluid under actual reactor 

operating conditions. The facility was sited just east of the Central Facilities Area. The OMRE 

facility included no provisions for the generation of electrical power and the heat produced was 

dissipated to the atmosphere using air-cooled heat exchangers. During the period from 1955 to 1958, 

testing on circulating organic fluids was performed with the most extensive reactor loop experiments 

occurring at the Materials Testing Reactor. Experiments that did not involve reactor loops were 

primarily focused on the measurement of heat transfer and the hydrodynamic characteristics of 

organic coolants. This work was done both at Atomics International and Aerojet-General.

The first OMRE criticality occurred on September 17, 1957. The reactor operated almost 

continuously until 1959, except for a period of several months near the end of 1958, when a new core 

was installed. Considerable information was obtained from OMRE on organic coolant stability and 

heat transfer conditions in the core. The operation of the reactor demonstrated the expected low 

organic radiation decomposition rates, low induced radioactivity, low corrosion rates, and the 

technical feasibility of designing and operating organic-cooled reactor plants. In all respects, the plant 

behaved as anticipated without occurrence of any unforeseen problems (AEC Annual Report, 1960). 

On November 16, 1960, an experiment was conducted to determine the feasibility of open-air burning 

of contaminated solvents that had accumulated at the OMRE facility. Approximately 400 gallons of 

liquid composed of diesel oil, xylene, methyl-chloroform, and a small amount of water were placed in 

an open vessel and ignited. 

OMRE was phased out in April 1, 1963 after accomplishing its intended purpose. Due to the minimal 

cost allowed for the construction of OMRE, the Experimental Organic Cooled Reactor (EOCR) was 

designed to extend and advance the OMRE studies (EOCR, 1959); however, the project was halted 

near the end of construction in December 1962 when the AEC determined that an organic-cooled 

reactor was not a significant improvement over other types of nuclear power reactors. The EOCR was 

sited near OMRE. One successful result of OMRE program was the municipal power reactor at 

Piqua, Ohio that operated from 1963 to 1966 (AEC, Jan-Dec1967). 

 
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5.1.4.3 Special Power Excursion Reactor Tests (SPERT) 

Since 1955, the vast majority of the AEC’s program for investigating the safety of water-cooled 

reactors has been centered at INL, when the Phillips Petroleum Company began reactor safety studies 

in the SPERT-1 reactor. Operating responsibility for the safety program was transferred to Idaho 

Nuclear Corporation in 1969 and reassigned to the Aerojet Nuclear Company in July 1970. These 

reactors were designed to operate at very high peak power for short periods of time. The SPERT tests 

contributed to the development and better understanding of engineering reactor safety control systems 

(NRTS Facilities, 1973). 

Early research centered on understanding the behavior and consequences of the reactivity (runaway 

power) accidents (SPERT-I, 1961). Four SPERT reactors were constructed and tested in an effort to 

understand the mechanisms that resist power increases and terminate runaway power conditions. 

These reactors were designed and operated beyond their normal safety limits to study a wide range of 

variables, such as core configuration, coolant flow and reflection, moderation, void, and temperature 

coefficients. Through 1962, most SPERT tests were performed with reactor cores of plate-type fuels, 

in which the highly-enriched uranium was alloyed and clad with aluminum or stainless steel in thin 

plates. A test of rod-type, slightly-enriched uranium that is more like the fuel found in the nuclear 

power industry was carried out in SPERT I in 1961 (SPERT-I, -II, and –III, 1957). By mid-1964, the 

SPERT project had shifted almost entirely to rod-type fuels. The SPERT reactor experiments fell into 

four major classifications:

1. Static experiments to determine core characteristics, such as void and temperature coefficients

2. Step tests in which the system is suddenly made supercritical

3. Ramp tests in which power is increased at a constant rate

4. Stability tests that may involve either spontaneous or externally-induced oscillations 

These tests were performed under various conditions of temperature, pressure, and coolant flow on 

cores of differing designs (SPERT Program, 1957). All operations were conducted from a central 

control building located approximately one-half mile from each reactor (see Figure 5-39) (AEC, 1954-

1986). The vast majority of the SPERT tests were conducted well below the threshold of core damage. 

Fuel damage was sustained in only a few tests.
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Figure 5-39: SPERT Area

SPERT-I start-up was on June 11, 1955 with operations completed in 1964. It was a simple open-tank

(4-feet diameter/14 feet deep), light-water moderated and reflected reactor located in a pit below 

grade. A bridge across the shield tank supported the drive mechanisms for the five cadmium control 

rods, instrumentation systems, and a closed-circuit television that allowed control room operators to 

view the reactor in operation. The SPERT-I tests were characterized by sudden power spikes arrested 

by shutdown or the self-limiting tendencies of the reactor itself. No provisions for heat removal or 

coolant circulation through the core were included. In the roughly ten years of operations, 

approximately 1300 kinetic tests were performed with six different reactor cores. After 1962, the 

primary emphasis was placed on extending the exploratory transient series of tests into a series of 

destructive transient tests. In1954, a 2.6 millisecond excursion at the Boiling Water Reactor 

Experiment, BORAX-I, produced an explosion that completely destroyed the reactor. Thus, there was 

considerable interest in studying the period region up to 2.6 milliseconds, which at that time was 

relatively unexplored (SPERT Program, 1957). The reason for this change in testing at SPERT-I was 

that other SPERT reactors (i.e., SPERT-IV) became operational. Prior to that time, care had been 

taken not to do serious damage to SPERT-I. Significant events at SPERT-I included: 

• A deliberate 2,300-megawatt power burst on November 5, 1962 that destroyed the highly-enriched 

uranium metal plate core and distorted the reactor vessel. This destructive test provided valuable 

data for simplification of safety design features. SPERT-I was rebuilt and used low-enriched 

uranium rod fuel afterward. 
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• A deliberate 17,400-megawatt power burst on November 12, 1963 using a low-enrichment (4% 

U-235) core of uranium oxide pins that successfully withstood the test without causing expulsion 

of water from the tank or damage to the reactor.

• A deliberate 35,000-megawatt power burst on April 14, 1964 (4% U-235 core) to demonstrate the 

damage-resistant capabilities of uranium oxide fuel pins, and that the failure of just a few pins 

would reduce the expected energy release, thereby protecting the rest of the core (NRTS Facilities, 

1973).

The November 12, 1963 and April 14, 1964 tests were major safety milestones in the nuclear power 

industry that employs low-enrichment uranium oxide fuel in light-water-cooled rectors. Figure 5-40 

shows a photo of the SPERT-I facility; note the instrument bunker under the soil berm on the side of 

the reactor building. 

Figure 5-40: The SPERT-I Facility 

SPERT-II start-up was on March 11, 1960 with operations completed in 1964. Unlike SPERT-I, it 

consisted of a closed, pressured water reactor with a coolant flow system designed for operation with 

either light or heavy water. It also had a much larger tank (10 feet diameter/24.5 feet deep). 

SPERT-II was not designed to operate at high temperatures and pressures; 375 psi and 400° F were 

upper limits. Flow rates up to 20,000 gallons per minute were achievable with SPERT-II. Coolant 

flow could be directed either upward or downward through the reactor core at velocities up to 25 feet 

per second to establish desired initial conditions prior to transient testing of the reactor (SPERT-I, -II, 

and –III, 1957). 
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SPERT-III start-up was December 19, 1958 with operations completed in 1968. The reactor was 

originally planned as the third in the series of SPERT reactors but was the second to be constructed. It 

was considered to be the most versatile facility yet developed for studying the inherent safety 

characteristics of nuclear reactors because it was designed for high pressures and temperatures. It 

provided the widest practical range of control over temperature, pressure, and coolant flow. Pressures 

from atmospheric to 2,500 psi, water temperatures from 68° to 670°F, and coolant flow rates ranging 

from zero to 20,000 gpm with heat-removal capacities up to 60,000 kilowatts for durations of 30 

minutes were attainable with SPERT-III (SPERT-III, 1961). 

SPERT-IV start-up was July 24, 1962 with operations completed in 1970. Unlike the other SPERT 

reactors, it was a twin-pool facility (open-tank) for detailed studies of reactor stability as affected by 

varying conditions, including forced coolant flow, variable height of water above the core, hydrostatic 

head, and other hydrodynamic effects. Most specifically, SPERT-IV made detailed studies of 

phenomena of instability first observed in the 1,300 excursions conducted at SPERT-I. The small size 

of the SPERT-I tank necessitated construction of the larger SPERT-IV facility in order to pursue 

further investigations of instability phenomena. Of particular interest was continuing the investigation 

of power instability under certain conditions during which there was boiling water in the core

(SPERT-IV, 1962). 

SPERT-IV was modified with a Capsule Driver Core to allow fuel samples to be inserted into a test 

hole in the center of the reactor core. The modified core accommodated various experimental 

capsules that contained fuels for transient response testing. The Capsule Driver Core operated on a 

transient basis with transient bursts lasting on the order of milliseconds. It did not accumulate large 

inventories of long-lived fission products. The total energy released per test was usually less than 300 

megawatts. The Capsule Driver Core operated for several years in SPERT-V to gain information on 

fuel destructive mechanisms pending completion of the Power Burst Facility (SPERT Tests, 1965). 

Construction of the Power Burst Facility was to extend the reactor safety studies at SPERT-I; start-up 

was not until September 1972, which is outside the period under evaluation in this report. 

5.1.5 Central Facilities Area (CFA)

ATTRIBUTION: Section 5.1.5 was completed by Mitch Findley, MJW Corporation. All conclusions 

drawn from the data regarding the Central Facilities Area were peer-reviewed by the individuals 

listed on the cover page. The rationale for all conclusions in subsequent sections of this document 

that are devoted to the Central Facilities Area are explained in the associated text.

Beginning in 1950, the Naval Base used during World War II as a site for testing guns reconditioned 

at the Naval Ordinance Plant in Pocatello, Idaho was converted to the Central Facilities Area (CFA). 

The old housing units were converted into offices that were used by the AEC and various AEC 

contractors. CFA served as the primary service and support center for INL programs (Stacy, 2000). 

While there were many centralized functions at CFA, there were few buildings and areas within CFA 

with any radioactive material hazards associated with them. Figure 5-41 shows the CFA in 1961 with 

buildings of interest designated. 
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Figure 5-41: Map of the Central Facilities Area in 1961 with Buildings of Interest

The following CFA buildings had hazards associated with them: 

• CF-610 and CF-616: Headquarters of the Idaho Nuclear Corporation’s Health and Safety Branch 

Manager, Industrial Safety Director, their staffs, and the CFA Health Physics Group, which was 

responsible for the CFA and certain associated areas. The radiation exposure records for all 

employees and visitors were maintained in this building. 

• CF-616: Headquarters of the INL respiratory protection and equipment program. 

• CF-669: The Central Facilities Laundry (constructed in 1950) laundered contaminated clothing or 

protective clothing that was used in radioactive areas. Laundry services included washing, drying, 

pressing, folding, and repairing of protective clothing. Because of the risk of possible 

contamination, no other items were washed at this laundry; non-contaminated clothing was sent to 

a commercial laundry. Clothing being sent to the CFA Laundry was segregated as to type (shoe 

covers, coveralls, gloves, etc.) and contamination level. After it was laundered and dried, the 

clothing was monitored for contamination on special tables. Each general type of clothing had its 

own permissible contamination limits. Any item that exceeded the recommended permissible 

limits was rewashed. After a defined decay period, Anti-C clothing that could not be 

decontaminated below those levels was bagged, labeled, and returned to the plant for disposal or 
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use as determined by area health physicists. The laundry had a radiation detector above the 

receiving room door and a continuous air monitor (CAM) in the working area. The CFA laundry 

was also used as an Anti-C change room and personnel decontamination facility for CFA 

personnel. 

• CF-674: The Sewage Treatment Plant, located near the laundry, processed any radioactive 

material placed in a drain at CFA. Most of the radioactive waste came from the Laundry 

(CF-669), although small amounts entered from CF-656 and CF-690. A hazardous area was 

created at the drying pond into which sewage was dumped. The dried sludge or sewage was 

removed periodically and shipped to the Burial Ground as a routine radioactive shipment. 

• CF-640 and CF-665: The Machine Shop and the Maintenance Shop may have had radioactive 

material in them from time-to-time. Radioactive material that the plant shops could not handle 

was worked on in the more fully-equipped CFA Machine Shop. Usually this material was of a 

low radiation and contamination level. The Maintenance Shop personnel would sometimes work 

on vehicles and equipment that were used to haul radioactive material. Such vehicles were 

surveyed prior to shop maintenance work and, when necessary, sent to CPP for decontamination. 

• CF-674A: The Chemical Engineering Laboratory was located at the south end of the CF-674 

Warehouse. Experiments involving radioactive tracers were prepared here; the resulting data was 

used for processing design at CPP. 

• CF-601: The CF Warehouse was the receiving center for radioactive shipments received via 

commercial carriers during regular working days. These shipments were monitored by CF Health 

Physics and the proper papers were completed and routed to the proper recipients (Savignac, 

1961).

The following buildings usually did not have hazards associated with them but are presented and 

described because of their importance to CF health physicists.

• Powder Bunkers: The powder bunkers were used as a temporary storage area for small lots of 

radioactive material sent to INL for burial. 

• Gantry – Rail Yard Area: All incoming or outgoing radioactive rail shipments, including 

piggyback trailer trucks, were routed through this area (see Figure 5-42). 
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Figure 5-42: ETR Reactor Vessel Arriving at CFA Gantry Crane in 1956

• CF-603: The CF Dispensary provided major medical services. The dispensary was completely 

staffed and equipped to offer NRTS contractors services ranging from physical examinations to 

evaluation and treatment of occupational injuries or minor non-occupational illnesses. The CFA 

dispensary also served as a bioassay sample collection station. In addition, a thyroid counter was 

located there.

• CF-690: AEC Health and Safety Technical Services included the External Dosimetry Branch, 

Analytical Chemistry Branch, Health Physics Branch, Instrumentation Branch, and Waste 

Management Branch. The Radiological and Environmental Sciences Laboratory provided 

radiobioassay and dosimetry services to INL until 1988. 

Associated Areas of CFA responsibility included:

• Burial Ground: Operational and health physics responsibility

• ARA Hot Cells: Health physics responsibility after 1965

• Site Vehicle Survey: Radiological survey of approximately 900 vehicles plus NRTS buses at set 

frequencies or as needed
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5.1.6 Burial Ground

ATTRIBUTION: Section 5.1.6 was completed by Mitch Findley, MJW Corporation. All conclusions 

drawn from the data regarding the Burial Ground were peer-reviewed by the individuals listed on the 

cover page. The rationale for all conclusions in subsequent sections of this document that are devoted 

to the Burial Ground are explained in the associated text.

The AEC recognized the need to have a local disposal ground for the solid radioactive waste 

generated during INL operations. In 1949-50, the U.S. Geological Survey (USGS) conducted 

extensive environmental monitoring in search of an appropriate burial site. In 1951, the site was 

selected due to the geology and hydrology characteristics of the land (see Figure 5-43). In May 1952, 

the first 13 acres were fenced in for shallow disposal of solid low-level waste. On July 3, 1952, the 

first trench was opened for the disposal of mixed-fission product (MFP) waste (see Figure 5-44) 

(RWMC History, 1979). Operation of the Burial Ground (also known as the Subsurface Disposal 

Area) was the responsibility of the Site Survey Branch of the AEC with waste disposal under the 

supervision of a health physicist of the Site Survey Branch. 

Figure 5-43: Road to the Burial Ground in 1961

Between 1952 and 1957, Trenches 1-10 were excavated. Metal tags on the fence enclosure marked 

trench locations. From the inception of the Burial Ground, it was policy to have a health physicist 

present to dump waste (RWMC History, 1985). 
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Figure 5-44: Burial Ground, circa 1955

On April 22, 1954, the Burial Ground received the first shipment of transuranic (TRU) waste from 

Rocky Flats. These shipments were to last until 1971, with Rocky Flats being the dominant 

contributor of TRU waste. Between April 1954 and November 1957, Rocky Flats TRU-contaminated 

waste was mixed with INL MPF-contaminated waste in Trenches 1 through 10. Rocky Flats waste 

arrived in the CFA by railcar where it was transferred to a flatbed truck and taken to the Burial 

Ground. The 55-gallon drums of Rocky Flats TRU waste were hand-stacked until 1963. Heavy items 

were emplaced with cranes. Normal operating practice was to include classified waste containers with 

other waste containers in the same trailer load. Shipments did not contain papers describing physical 

form and radionuclide content until 1964. Rocky Flats provided a memo at end of each year with a 

summary of the total radionuclide content and volume shipped to INL (RWMC History, 1979). 

In 1957, the Burial Ground was expanded from its initial 13 acres to 88 acres. At that time, a routine 

twice-weekly pick-up of site-generated waste was initiated. The year 1957 also marked the 

excavation of the first burial pits for the receipt of bulky items from Rocky Flats as well as the 

transition of burial of Rocky Flats waste drums from trenches to pits (see Figure 5-45) (Remediation, 

2006). 
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Figure 5-45: Emplacement of Rocky Flats Drums in a Burial Ground Pit in 1958

In May 1960, the AEC designated the INL Burial Ground, and one at ORNL, as interim burial 

grounds for waste from other AEC sites as well as non-AEC waste generators. As a result of the 

AEC’s decision, large quantities of beta-gamma-contaminated waste were received from private 

industry, universities, the U.S. Armed Forces, and AEC contractors. INL and ORNL established 

formal policies and procedures for dealing with improper or problem shipments. On May 21, 1961, a 

separate burial ground opened near SL-1. The burial ground included one trench and two pits for 

contaminated materials from the January 3, 1961 SL-1 accident so that waste would not have to be 

transported over 16 miles away to the Burial Ground (RWMC History, 1985). 

On October 1, 1962, the responsibility for managing and operating the Burial Ground was transferred 

from the Site Survey Branch of the AEC to Phillips Petroleum Company, which had been acting as 

AEC’s agent in operating the Burial Ground (RWMC History, 1985). In June 1963, the AEC 

announced its decision to withdraw its services as an interim burial ground for radioactive waste 

effective August 12, 1963. The decision was made because other suitable burial sites had been 

established by private industry. Rocky Flats waste was still received at INL because privatelyoperated burial grounds were not allowed to receive classified or sensitive waste. In November 1963, 

a decision was made to no longer stack Rocky Flats waste, but rather, to dump the drums into the 

burial pits to reduce labor costs and minimize personnel external exposures (see Figure 5-46). This 

practice lasted until late 1969 (RWMC History, 1979). 
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Figure 5-46: Mass Dumping of Rocky Flats TRU Drums in 1969

Beginning in April 1964, waste shipments from Rocky Flats were declassified and no classified 

materials were to be included in future shipments to the INL Burial Ground. Also, around that time, 

the first shipments of waste from the Coors Porcelain Company (via Rocky Flats) began arriving at 

INL. These shipments consisted of thirty 55-gallon drums of low-level contaminated waste from the 

Pluto Program that were generated between 1964 and 1970 (RFP Waste, 2005). 

In January 1969, after two days of rainfall and thawing ground, water runoff filled Pit 10 and much of 

Trenches 48 and 49 (see Figure 5-47). Snowdrifts blocked the existing drainage that was installed 

after the 1962 “Chinook” flood. The runoff topped old dikes and flowed through the Burial Ground 

causing heaving of buried waste. Later, the dikes would be raised and the exterior drainage ditches 

enlarged so heavy equipment could clear snowdrifts. A major fire at Rocky Flats in Building 776 on 

May 11, 1969 was the initiating event for major changes to come for the Burial Ground. The two-year 

clean-up from the fire dramatically increased waste shipments to INL and caused Idaho Senator Frank 

Church to request four federal agencies, including the Bureau of Radiological Health, to conduct a 

joint review of the Burial Ground. As a result, the AEC recommended suspension of the burial of 

intermediate-level waste from Rocky Flats during the winter. It was also noted that experience at 

other facilities suggested that segregation of plutonium-contaminated waste was advisable (RWMC 

History, 1985). 
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 Figure 5-47: January 1969 Flooding of Burial Ground 

In late 1969, there was a retrieval effort in Pit 1 for some experimental equipment that was believed to 

have been inadvertently buried. It marked the first time in the history of the Burial Ground that waste 

was deliberately exhumed; it foreshadowed future remediation work that is now the primary focus of 

work currently performed at the facility (RWMC History, 1985). 

During 1970, several actions resulted in discontinuing burial of plutonium-contaminated waste at INL. 

In January, a directive from the AEC-ID Manager stopped burial of Rocky Flats waste at the Burial 

Ground during winter and spring, and required that Rocky Flats waste be stacked above ground. On 

March 20, 1970, the AEC issued Immediate Action Directive No. 0511-21 requiring segregation of all 

waste contaminated with long-lived TRU waste > 10 nCi/g. In October 1970, the construction of 

Transuranic Storage Area-1 (TSA, aka Idaho Transuranic Storage Area) was completed and the first 

waste was stored on TSA-1 on November 9, 1970. In 1971, the first buildings were installed at the 

Burial Ground, consisting of a trailer for Burial Ground operations and Health Physics, and a TSA 

clothing-change trailer. The installation of these trailers is outside the period under evaluation for this 

evaluation for this report (Remediation, 2006). 

Figure 5-48 shows a schematic of a Transuranic Storage Area. Figure 5-49 shows a current map of 

the Burial Ground and other facilities associated with the Radioactive Waste Management Complex. 
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Figure 5-48: Schematic of a Transuranic Storage Area

Figure 5-49: Current Map of the Burial Ground and Associated Facilities
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5.2 Internal Radiological Exposure Sources from INL Operations

Radionuclides with the widest application throughout INL facilities during the period under 

evaluation are: 

• Fission and activation products (e.g., Sr-90, Cs-137, Zn-65, Co-60) 

• Exotic radionuclides (e.g., C-14, Ca-45, P-32, S-35, Ba-140/La-140, and others produced by 

reactor neutron irradiation, including transuranic radionuclides) 

• Plutonium (Pu-238, Pu-239, Pu-240, Pu-241) 

• Uranium (U-238, U-233, U-234, U-235, U-236) 

• Thorium (Th-232) 

The following subsections provide an overview of the internal exposure sources specific to the six 

operating areas for the INL class under evaluation. 

• Test Reactor Area (TRA) – Section 5.2.1 

• Chemical Processing Plant (CPP) – Section 5.2.2 

• Test Area North (TAN) – Section 5.2.3 

• Misc. Reactor Areas (SPERT, ARA, GCRE, OMRE) – Section 5.2.4 

• Central Facilities Area (CFA) – Section 5.2.5 

• Burial Ground – Section 5.2.6 

5.2.1 Internal Radiological Exposure Sources from Test Reactor Area Operations

The diversity of operations at the Test Reactor Area makes it difficult for NIOSH to collect complete 

details of all TRA internal exposure sources. Nevertheless, NIOSH has determined that the internal 

dose potential during the period under evaluation is associated with reactors, laboratories, machine 

shops, and other research and support activities as well as radioactive waste disposal. Although not 

all specific areas where waste materials were generated and buried are readily identifiable from the 

documents available to NIOSH, a potential for internal intakes from these activities is assumed to 

exist. The following radionuclides specific to TRA operations are discussed in the subsections below 

and in Section 7.2.: 

• Fission and activation products (e.g., Sr-90, Cs-137, Zn-65, Co-60) 

• Exotic radionuclides (e.g., C-14, Ca-45, P-32, S-35, Ba-140/La-140, and others produced by 

reactor neutron irradiation, including transuranic radionuclides) 

• Plutonium (Pu-238, Pu-239, Pu-240, Pu-241) 

• Uranium (U-238, U-233, U-234, U-235, U-236) 

• Thorium (Th-232 and Th-228/Th-230) 
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5.2.1.1 Fission and Activation Products

Since the inception of the MTR in 1952, fission and activation product exposure potential has existed 

at TRA. This exposure potential continues to the present because the ATR is still in operation. Thus, 

potential exposure to fission and activation products existed throughout the period under evaluation 

(1949-1970), as indicated in surveys, air samples, and bioassay samples for fission and activation 

products (via gross alpha and gross beta/gamma analysis) (INEL Bioassay, 2010a; INEL Bioassay, 

2010b).

5.2.1.2 Exotic Radionuclides

TRA reactors used neutron irradiation to generate radionuclides beyond the spectra of fission and 

activation products (e.g., C-14, Ca-45, P-32, Ba-140/La-140, and others). NIOSH has used 

documentation, data, and shipping receipts for these operations to develop a list of radionuclides 

handled at TRA during the period under evaluation. NIOSH has not captured shipping reports for 

some years, so this list is not considered a complete inventory. The list includes fission products, 

activation products, and actinides, including transuranics. 

Experiments (and their resulting radionuclides) were transferred to and from the reactors using the 

sample rabbits, pneumatic transfer tubes, and discharge chutes mentioned in Section 5.1.1 as well as 

other shielded apparatus. Some testing of experiments was conducted directly from fuel bins while 

other testing was performed in reactor canals. After some cooling in reactor canals, some experiments 

were repackaged directly to shipping casks and returned to the project requestor. Samples were taken 

from some experiments after transfer by pneumatic tubes to either hot cells or the alpha cave. Those 

samples were transferred to either the Fast Chopper for cross-sectional analysis or to one of the labs at 

TRA or CPP via a specially-designed, shielded carrier. NIOSH has not found documentation to show 

that TRA personnel handled irradiated or highly-radioactive samples in open lab environments. 

NIOSH has captured documentation of incidents involving loss of containment of radioactive 

materials. Such incidents are discussed in Section 5.4.1. 

Small amounts of tritium were generated through research that entailed highly irradiating beryllium, 

once in 1960 and once in 1961. Gases produced by the irradiations were analyzed in a vacuum system 

housed in a fume hood that was originally in the lab area and then moved to one of the alpha labs in 

1961 (Mahathy, 2015). The beryllium gas was melted and handled in the vacuum system with 

helium, hydrogen, and tritium components separated by a palladium trap. The measured yield of 

tritium gas inside the vacuum system ranged from 2.7 to about five percent of the gas (Mahathy, 

2015). 

5.2.1.3 Plutonium

Potential for plutonium exposure existed at TRA during the entire period under evaluation, and at each 

of the primary reactors (MTR, ETR and ATR). Plutonium fuel cores were used twice at MTR, for one 

cycle in 1958 and for about five months in 1970. The latter fuel core was disassembled at the ARA 

Hot Cell in December 1970 (Mandiloff, 1971). Plutonium fuel plates and assemblies received from 

other sites, including Hanford, ORNL, and University of California Radiation Lab were irradiated at 

MTR through much of its operating history. After irradiation, these materials, sometimes cooled in 

the MTR canal and with some analyses, were returned to the originator. Examples of plutonium 
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materials shipped back post-irradiation are: (1) One irradiated plutonium-aluminum alloy 

"napkin-ring" shipped to Argonne National Lab in January 1954 (MTR Progress Report, 1953); (2) 

plutonium in metallic form shipped to Lawrence Radiation Lab in October 1962 (Bengstad, 1962, 

PDF p. 17); and (3) plutonium in metallic form shipped to Lawrence Radiation Lab in May 1963 

(McKenna, 1963, PDF p. 19). 

TRA work involving plutonium is supported by the existence of routine bioassay as well as 

incident/air sample-based plutonium bioassay data for TRA workers. With the exception of work 

performed in the Alpha Labs for which incident-based plutonium bioassay data are available, all other 

potential plutonium exposures would have included concurrent exposure to mixed fission products for 

which there was routine bioassay.

5.2.1.4 Uranium

The potential for uranium exposure existed at TRA during the entire period under evaluation but the 

potential was very small because TRA received encapsulated materials, irradiated them, and either 

stored them as spent fuel or returned them to the CPP - without opening the capsules. On some 

occasions, MTR lab staff analyzed samples made of U-233 in hot cells. The MTR metallurgical lab 

also fabricated test fuel capsules using depleted and enriched uranium. Uranium would include U238, U-233, U-234, U-235, and U-236. 

5.2.1.5 Thorium

The potential for thorium exposure existed at TRA during the entire period under evaluation. Many 

irradiations were performed using encapsulated materials to test the nuclear and physical properties of 

thorium for use as a fuel. Samples of such irradiated thorium were collected in the reactor and canal 

but transferred in shielded rabbits using pneumatic tubes. Some thorium was irradiated to make 

U-233 and Pa-233. In both cases, the end products were transferred and handled in shielded 

containment. Figure 5-21 shows a worker preparing pure Pa-233 from irradiated thorium. 

5.2.2 Internal Radiological Exposure Sources from CPP Operations

The primary function of the Chemical Processing Plant was to process spent fuel elements containing 

enriched uranium for the recovery of the un-fissioned uranium. Fuel elements were dissolved in acid 

and the uranium was separated from fission products by continuous liquid-liquid extraction, 

employing tributyl-phosphate (TBP) and methyl iso-butyl ketone (MIBK) as solvents (Analytical Lab, 

1957). 

Much of the processing equipment in the process building was located in heavily-shielded cells and 

had to be operated remotely, reducing the chances for an internal radiological exposure. However, the 

entire plant design was based on a direct-contact maintenance philosophy (i.e., maintenance was 

performed by direct manual contact only during periodic shutdowns and then only after the particular 

cell and equipment were thoroughly decontaminated) (CPP Safety Review, 1974). In addition, during 

processing, samples were taken of the dissolver solution, the raffinates from all three cycle extraction 

columns, and the final product solution (Analytical Program, 1957).
 

 

 

 

 

 

 

 

 

 

 

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Radionuclides with the widest application throughout CPP facilities during the period under 

evaluation are listed below. They are discussed in the following subsections and in Section 7.2. 

• Uranium 

• Neptunium 

• Plutonium 

• Fission and Activation Products 

• Barium/Lanthanum 

• Thorium 

• Radioactive Noble Gases 

• Iodine 

• Tritium 

NOTE: When radioactive noble gases are released into the atmosphere, external exposures dominate. 

Whole-body doses resulting from immersion in rare gases are usually calculated by considering only 

external irradiation. The results ignore any contributions to dose resulting from the absorption of 

inhaled gas in the blood and its subsequent diffusion into other tissues. However, radioactive noble 

gases are not entirely insoluble in body fluids. Hence, for a person immersed in a cloud of noble gas, 

the whole-body dose from the gas absorbed by the body tissues following inhalation should be 

considered in addition to the dose from the radiation external to the body. 

Radiation doses resulting from the inhalation of noble gases can be estimated using the dose 

conversion factors shown in Table 5-6, where all values are in units of mrem per one minute of 

submersion in a concentration of 1 μCi/g (air): 

Table 5-6: Dose Conversion Factors for Kr-85 and Xe-133 

Organ Radionuclide 

Kr-85 Xe-133 

Tracheal Mucosa 0.086 0.116 

Lung 0.032 0.020 

Fatty Tissues 0.005 0.011 

Lean Tissues 0.0006 0.0019 

Gonads 0.0006 0.0012 

This same krypton concentration and exposure time would produce a total external dose of 3.847 

mrem to the skin. 

Because radioactive noble gases do not represent a significant internal exposure concern for INL, 

NIOSH finds there is no need to assess the internal doses attributable to radioactive noble gases. 

5.2.2.1 Uranium 

The function of the Idaho Chemical Processing Plant was to process fuel elements containing uranium 

enriched in U- 235 for recovery of the un-fissioned uranium. The uranium isotopes in feed samples 

were U-234, U-235, U-236, and U-238. Routine samples were taken of the dissolver solution, the 

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aqueous raffinates from all three cycle extraction columns, and the final uranyl nitrate product. The 

high concentration of fission products in the dissolver solution and first-cycle extraction raffinate 

required the use of remote-handling facilities for separation and analysis. In addition, the high activity 

of the raffinates required that samples be limited to 1 mL or less. Uranium sampled from column 

raffinates usually contained 1 x 10-3 mg uranium per mL of sample (Analytical Program, 1957). 

In 1957, two thorium slugs that had been irradiated in the MTR were dissolved and analyzed for total 

U-233. Each slug had a total mass of 1730 g, with 78 grams being aluminum cladding. Separation of 

the uranium was performed in a glove box and small samples were transferred to the Mass Lab for 

analysis (Uranium, 1957). In addition, U-233 was used as a spike isotope in the analytical 

laboratories (Analytical Program, 1957). 

U-233 and U-232 are formed by the following series of reactions:

5.2.2.2 Neptunium

Neptunium-237, an alpha-emitting radionuclide with a half-life of 2.1 million years, tends to 

concentrate on the bone surfaces when inhaled or ingested. The isotope is produced in uranium-fueled 

reactors by one of the two following reactions: 

Which of the two reactions dominates the process depends on the degree of fuel enrichment. In 

natural uranium, approximately 70% of the neptunium is produced by the first reaction, and the 

remainder is produced by the second (Neptunium, 1972). The highly-enriched fuels processed at CPP 

would have seen neptunium produced primarily through the second reaction (Uranium, 1966). 

In 1958, the CPP Technical Division requested the development of a method for determining 

neptunium in process-stream samples. Analytical work done in support of this request was carried out 

in the Remote Analytical Facility, along with work investigating the possibility of recovering Np-237 

from process raffinates (CPP Progress Report, Jan1958). A satisfactory method for determining the 

concentration of neptunium in process streams was documented in February, 1958 (CPP Progress 

Report, Feb1958). A simplified process was recorded in July 1958 (CPP Progress Report, Jul1958), 

and routine analysis of plant process streams for Np-237 concentrations began in August 1958 (CPP 

Progress Report, Aug1958). Due to operational difficulties in other plant processes, the test program 

to concentrate neptunium from plant streams was abandoned in September 1958 (CPP Progress 

Report, Sep1958). A second, short-lived pilot program was initiated in May 1959 and continued for a 

few months (CPP Progress Report, Jun1959).
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Because of the differing chemistry between uranium and neptunium, most of the neptunium in the 

reactor fuels processed prior to 1965 was lost to first-cycle raffinate. The feasibility of recovering 

Np-237 from reactor fuels at CPP was demonstrated during a 1964 aluminum-alloyed fuel-processing 

run by recovering over 90 percent of the neptunium in the second-cycle raffinate generated during the 

campaign (i.e., 18.8 g of the estimated 32 g available) (Uranium, 1966). Following these preliminary 

demonstration tests, routine recovery of Np-237 at CPP began in 1965. In the separation scheme, 

Np(VI) was co-extracted with uranium and some plutonium into the first-cycle 5% TBP solvent. 

Neptunium and plutonium accompanied the uranium through the first-cycle scrubbing and stripping 

columns. Partitioning of uranium from neptunium and plutonium was then effected in the secondcycle acid-deficient hexone-extraction column by addition of ferrous sulfamate to reject neptunium 

and plutonium to the raffinate stream.

The accumulated raffinate stream was then processed in a separate operation for neptunium recovery 

and concentration by evaporation. The concentrated neptunium product obtained at this point was 

essentially free from fission products but still contained plutonium and some uranium (Neptunium, 

1972). In January 1969, the accumulated solution (containing 1.9 kg of Np-237) was sampled in J cell 

and further concentrated to about 250 L before being stored in a vessel in Cell N as an impure nitric 

acid solution (Neptunium, 1969).

In total, these raffinate streams were collected over a seven-year period. Final processing of the 

Np-Pu-U solution was successfully completed during a three-week campaign in June 1972. The 

uranium was partitioned from the neptunium and plutonium, then the neptunium and plutonium 

solution was decontaminated from fission products. The final solution, containing 5412 g of 

neptunium and 544 g of plutonium, was shipped off-site for eventual neptunium purification 

(Neptunium, 1973). 

5.2.2.3 Plutonium

Plutonium would include Pu-238, Pu-239, Pu-240, and Pu-241. On a mass basis, Pu-239 is usually 

the most prevalent isotope of plutonium in spent nuclear fuels. However, because of the relativelyshort half-life of Pu-238 (88 years) in most fuels, more than 90% of the plutonium activity is due to 

the presence of Pu-238. Most of the fission products were extracted from the process solution into the 

1A raffinate stream. The plutonium activity may or may not have followed the fission products, 

depending upon the nature of the fuel. In general, the Pu-238 activity became the dominant 

radiological hazard once the process made it past the 1A column. The first-cycle product was highly 

concentrated and presented a significant plutonium hazard, which was decontaminated from fission 

products. 

Plutonium continued to dominate the potential internal radiological hazards throughout the second and 

third process cycles until the product proceeded beyond the 3AR stream. At that point, the U-234 

activity became the dominant hazard (CPP Dosimetry, 1984).

Highly-enriched U-235 fuels were processed at the CPP to recover residual U-235. Because initial 

processing rates were low, the plant was designed using essentially pilot-plant principles, and 

maintenance was designed to be performed by cell entry and contact techniques. Due to the 

relatively-low U-238 content in the fuel, plutonium was judged not to be a problem. However, in the 

course of operation, the burn-up of fuels increased; and in some fuels, the U-238 content was higher 
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than originally expected. This led to questions about whether plutonium might be a current problem, 

or might become a problem with predicted higher burn-ups (Plutonium, 1980).

For most of the CPP operating period, plutonium was regarded as an undesirable contaminant to the 

final uranium product. In 1954, a sample was taken from the final product solution and analyzed for 

plutonium contamination and found to contain less than 1 part-per-billion plutonium (CPP Progress 

Report, Feb1954). In November 1957, CPP Engineering Design Section was requested to prepare 

plans and a cost estimate for the conversion of X-Cell into a radiochemical development area capable 

of handling plutonium (CPP Progress Report, Nov1957). In January 1958, MTR Technical requested 

the analysis of four fuel elements from the MTR 20%-enriched-uranium fuel core. However, due to 

the lack of adequate plutonium-handling facilities, a complete analysis could not be made at the CPP 

(CPP Progress Report, Feb1958). As a workaround, the dissolution of the elements occurred in LCell, radiochemical separation was conducted in the Remote Analytical Facility, and mass analysis 

was conducted at Argonne (CPP Progress Report, Feb1958). 

Although plutonium would have been present in the process and waste streams, NIOSH has found no 

indication that partitive equipment was ever added to the plant that would allow the plutonium to be 

separated from the waste stream (Stroschein, 1967; CPP Phase-out, 1994). 

5.2.2.4 Fission and Activation Products

Fission and activation products present in spent fuel include Sr-90, Cs-137, Zn-65, Co-60 and 

numerous others. Over 99 percent of the fission products present in the spent fuel would exit the 

process in the first-cycle extraction raffinates. Dissolver solutions passed through several cycles of 

countercurrent solvent extraction to remove essentially all fission products and transuranic elements 

from the final product uranium solution. Prior to construction of the WCF, these wastes were 

segregated by fuel types and stored in either 30,000-gallon tanks designated for zirconium- and 

stainless-steel-clad fuel wastes, or 300,000-gallon tanks designated for aluminum-clad fuels. These 

wastes were packaged into 500-gallon lead-shielded spheres and shipped to the ORNL fission product 

recovery facility (CPP Waste Management, 1960). 

Other liquid wastes from process-equipment drains, cell-floor drains and decontamination-area drains, 

amounting to a million gallons per year, were collected in tanks before transfer to the process 

equipment waste evaporator for concentration. Table 5-7 presents a summary of the estimated 

high-level liquid radioactive waste volumes and the associated activity for fiscal years 1972 through 

1975. Between December 1963 and May 1974, over 2.7 million gallons of liquid wastes containing 

45.8 x 106 Ci of fission products and transuranics were sent to the WCF.
 

 

 

 

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Table 5-7: Fiscal Year-End Inventories of Radioactive Liquid Waste Activities 

Year Nuclide Activity (Ci) 

1972 

All* 20 X 106 

Sr-90 3 X 106 

Cs-137 4 X 106 

1973 

All 53 X 106 

Sr-90 6 X 106 

Cs-137 7 X 106 

1974 

All 78 X 106 

Sr-90 6 X 106 

Cs-137 7 X 106 

1975 

All 90 X 106 

Sr-90 9 X 106 

Cs-137 9 X 106 

Source: Waste Management, 1974 

* “All” radionuclides includes fission products as well as transuranics. 

Table 5-8 presents a summary of the estimated calcined solid waste volumes and the associated 

activity for fiscal years 1970 through 1975. 

Table 5-8: Fiscal Year-End Inventories of Calcined Solid Waste Activities 

Year Nuclide Activity (Ci) 

1970 

All 3.3 X 107 

Sr-90 7.1 X 106 

Cs-137 8.3 X 106 

1971 

All 3.4 X 107 

Sr-90 7.8 X 106 

Cs-137 9.0 X 106 

1972 

All 4.1 X 107 

Sr-90 9.1 X 106 

Cs-137 10.5 X 106 

1973 

All 4.8 X 107 

Sr-90 10.1 X 106 

Cs-137 11.6 X 106 

1974 

All 5.0 X 107 

Sr-90 10.5 X 106 

Cs-137 12.0 X 106 

1975 

All 6.8 X 107 

Sr-90 14.3 X 106 

Cs-137 16.2 X 106 

Source: Waste Management, 1974
	

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5.2.2.5 Barium/Lanthanum

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While most of the plant was concerned with the recovery of uranium from spent fuel and fission 

products, the RaLa process was concerned with the recovery of a fission product. RaLa stands for 

radioactive lanthanum, a fission product valued for its high-energy gamma radiation. 

In the RaLa process, a fresh fuel element just out of the MTR was dissolved in caustic, and Ba-l40 

extracted by a series of precipitations and re-dissolutions. Barium-l40 was then allowed to decay to 

La-l40 which then could be separated and used (Stroschein, 1967). 

5.2.2.6 Thorium

In 1957, two thorium slugs that had been irradiated in the MTR were dissolved and analyzed for total 

U-233. Each slug had a total mass of 1730 g, with 78 grams being aluminum cladding. Separation of 

the uranium was performed in a glove box and small samples were transferred to the Mass Lab for 

analysis (Uranium, 1957).

In 1961, the Analytical Section was requested by the Reactor Physics Research Group to analyze six 

irradiated thorium slugs for uranium concentration and isotope ratio, aluminum, thorium, Cs-137, and 

Zr-95. Each of the six slugs had a mass of about 1730 g and were dissolved and prepared for analysis 

in the Multicurie Cell. Manipulations within the cell were made with a pair of master-slave 

manipulators (Thorium, 1961).

5.2.2.7 Radioactive Noble Gases

The Rare Gas Plant was operated at CPP to recover krypton and xenon from gases released during the 

dissolution of reactor fuels. The plant was designed to recover several thousand curies per day of 

krypton and a significant volume of xenon from off-gas evolving from the acid dissolution of spent 

nuclear reactor fuel elements. Radioactive isotopes of interest liberated during dissolution are 

presented in Table 5-9. The original process used charcoal absorbers (operating at liquid nitrogen 

temperatures) for removing radioactive krypton from the dissolver off-gas. When the absorbent was 

saturated, it was heated to drive off the rare gases which were then collected in a cold trap. However, 

the product from this early process was impure, and the cooling requirements of the process required 

more nitrogen than the existing liquid nitrogen plant could produce on a continuous basis (Offutt, 

1969). 

Table 5-9: Radioactive Noble Gases Released During Dissolution of Spent Fuel

Isotope Half-life Emission

Kr-85 10.76 years 0.687 MeV βXe-133 5.243 days 0.427 MeV βXe-135 9.14 hours 1.165 MeV β-
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Initially, the plant only operated during the dissolution of aluminum-alloyed fuels due to concerns 

about the combustion potential of the hydrogen-rich off-gas from other fuel types (CPP Krypton, 

1979). The plant was not upgraded to allow processing of hydrogen-rich off-gas streams until 1987. 

As discussed in the introduction to Section 5.2.2, because radioactive noble gases do not represent a 

significant internal exposure concern, NIOSH finds there is no need to assess the internal doses 

attributable to radioactive noble gases. 

5.2.2.8 Iodine

The initial operating theory at CPP was that fuels processed were cooled sufficiently that radioiodines 

and short-lived noble gases should have decayed to insignificant levels (Waste Management, 1974). 

However, during December 1956, 495 fuel elements were dissolved at CPP with 175 of those 

elements having been discharged on September 11, 1956, and the remaining 420 elements having 

been discharged on September 29, 1956. This led to the release of I-131 quantities shown in Table 

5-10. 

Table 5-10: I-131 Release from C Cell Dissolving Operations During December 1956

Dates Curies of I-131

December 1-10 0

December 11-12 7.2

December 13 0

December 14-15 35.5

December 16-19 59.7

December 20-31 0

Because of a cooling time of only two days, compared to the 90-120 days cooling normally provided 

for most other fuel elements before reprocessing, the RaLa process contributes large quantities of 

fission product iodine to the process off-gas. Each fuel element used in this process contained 

approximately 200 grams of uranium when charged into the reactor. Processing of the fuel was 

accompanied by release of approximately 25,000 curies of I-131, which needed to be trapped and 

confined (CPP RaLa, 1969). Most of the iodine found in the vessel off-gas system was released 

during solution transfers, particularly transfer of the process wastes to storage. The use of' 

compressed air to inject reagent solution into the centrifuge bowl, and the consequent escape of 

iodine-contaminated gases from the centrifuge shaft seal, was the major cause of iodine release to the 

cell atmosphere and ventilation system (I-131 Containment, 1961).

The virtually complete containment of iodine was a major objective in the design of the RaLa process. 

The RaLa system was designed for a vacuum of 10 to 20 inches of water in all process equipment to 

ensure that any leakage would be from the cell atmosphere into the equipment. All the process 

off-gas, plus any air leakage into the system, was to be passed through an iodine-removal scrubber 

before release to the atmosphere through the 250-foot plant stack. However, the first processing of a 

fully-irradiated fuel element disclosed serious deficiencies in the iodine-containment measures. The 

peak iodine release rate to the process off-gas was considerably higher than had been anticipated. The 
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transfer of the dissolution supernate waste solution through the plant lines and vessels to permanent 

storage released about 200 curies of I-131 to the main CPP vessel off-gas system. 

An upgraded RaLa Off-Gas system contained charcoal beds for removing radioiodine and a gas holder 

for storing the noble gases during decay (Waste Management, 1974). This system was able to remove 

gaseous iodine with a 99.9% efficiency, but was still limited in its ability to remove iodine-bearing 

dusts and mists (I-131 Containment, 1961).

5.2.2.9 Tritium

From 1953 (when CPP operations began) to 1984, CPP wastewater containing tritium was routinely 

injected into the Snake River Plain aquifer through a 580-ft-deep disposal well. Beginning February 

9, 1984, the wastewater was routinely sent to an infiltration pond; a second pond was put into use on 

October 17, 1985. The disposal well was used as an emergency back-up to the disposal ponds from 

1984 to 1986. 

During the 1971 zirconium-processing campaign, several streams in the first-cycle uranium extraction 

system were sampled and analyzed for tritium. The results from five different sample sets were 

averaged and a tritium material balance was made using the tritium entering the system in the IFU 

stream as the base value. The fraction of tritium in the IBR, ICR, and ICP streams was found to be 

very small. The ICR organic stream is recycled to the extraction column as the IAX stream. This 

organic stream appeared to be holding a constant tritium concentration equivalent to about 0.2% of the 

total entering the system. The three mixer settlers did not remove any significant fraction of tritium 

from this stream. The IBR stream, with 1.8% of the tritium balance, was recycled to the extraction 

column as part of the IAF stream. The tritium left the system in the ICP and IAR streams. The ICP 

stream contained less than 0.1% of the incoming tritium; therefore, by difference, the IAR stream 

removed about 99.9% of the tritium from the system. 

Additional tritium sampling during a campaign in which aluminum-alloy and zirconium-alloy fuels 

were processed together indicated that, during dissolution of aluminum fuel in nitric acid, 

approximately 45% of the tritium was released with the dissolver off-gas and 55% entered the 

extraction system. During dissolution of zirconium fuel in hydrofluoric acid, approximately 90% of 

the tritium was released to the off-gas and 10% entered the extraction system (Chemical Programs, 

1972). 

5.2.3 Internal Radiological Exposure Sources from TAN Operations

Radiological operations at TAN have taken place since 1955. TAN was primarily a nuclear reactors 

operations area with no known radiological separations work, with the exception of the worked 

performed in the area’s radiochemistry laboratories. The following categories of radionuclides that 

were specific to TAN operations are discussed in the subsections below and in Section 7.2. 

• Mixed fission and activation products

• Radioiodines (I-129, I-131, I-132, I-133, I-134, and I-135) 

• Radioactive noble gases (Ar-41, Kr-85m, Kr-87, Kr-88, Xe-129m, Xe-135, Xe-135m, and 

Xe-138) 
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• Plutonium (Pu-238, Pu-239, Pu-240, Pu-241, Pu-242) 

• Neptunium (Np-235, Np-237) 

• Uranium (U-232, U-233, U-234, U-235, U-236, U-238) 

• Other Actinides (Ac-227, Th-228, Th-230, Th-232, Pa-231, Am-241, Am-243, Cm-242, Cm-243, 

Cm-244, Cm-245, Cm-246) 

• Other radionuclides (e.g., H-3, Be-10, C-14, Na-24, Al-28, Fe-55, Co-60, Zn-65, Sr-90, Zr-95, 

Nb-93, Nb-95, Ag-110, Sb-122, Sb-124, Cs-137, Cs-139, Ce-141, Ce-144, Pr-140, Tm-170, 

Hf-175, Hf-181, Ta-179, Ta-182, W-185, W-187, Tl-204, Po-210, Ra-226, and others produced by 

reactor neutron irradiation)

5.2.3.1 Mixed Fission and Activation Products

Mixed fission and activation products have been present at TAN since radiological operations started 

there in 1955. Fission and activation products were produced by the various reactors at TAN and 

were received at TAN via shipments of radioactive materials (e.g., irradiated fuels, contaminated 

equipment). Because radioiodines and radioactive noble gases were periodically present at TAN in 

significant quantities without any other fission and activation products being present, these fission 

products are addressed in their own subsections below. 

5.2.3.2 Radioiodines

At TAN, radioiodines (e.g., I-129, I-131, I-132, I-133, I-134, and I-135) were produced by the various 

reactors and with the exception of the IET Facility, appear to have only been released in small 

quantities. Large quantities of radioiodines were released to the atmosphere as episodic releases from 

operations at the IET Facility (DOE, 1991, PDF pp. 205-212). During the ANP Program, episodic 

releases of over 60,000 Ci of radioiodines occurred at the IET Facility. During the SNAPTRAN tests, 

episodic releases of radioiodines were less than 30 Ci at the IET Facility. Most of the ANP Program 

releases were via the IET Facility’s 150-foot-high stack. However, the releases from the SNAPTRAN 

tests were not vented via the IET Facility’s stack; they occurred at ground level (SNAPTRAN, 1964; 

SNAPTRAN, 1967).

5.2.3.3 Radioactive Noble Gases

At TAN, radioactive noble gases (e.g., Ar-41, Kr-85m, Kr-87, Kr-88, Xe-129m, Xe-135, Xe-135m, 

and Xe-138) were produced by the various reactors and, with the exception of the IET Facility, appear 

to have only been released in small quantities. Large quantities of radioactive noble gases were 

released to the atmosphere as episodic releases from operations at the IET Facility (DOE, 1991, PDF

pp. 205-212). During the ANP Program, episodic releases consisting of over 100,000 Ci of 

radioactive noble gases occurred at the IET Facility. During the SNAPTRAN tests, episodic releases 

of radioactive noble gases were less than 2,000 Ci at the IET Facility (DOE, 1991, PDF pp. 205-212; 

SPERT and SNAPTRAN, 1960s, PDF pp. 65-79 and 82-118). 

As discussed in the introduction to Section 5.2.2, because radioactive noble gases do not represent a 

significant internal exposure concern, NIOSH finds there is no need to assess the internal doses 

attributable to radioactive noble gases. 
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5.2.3.4 Plutonium

Plutonium has been present at TAN since radiological operations started there in 1955. Plutonium 

was produced by the various reactors at TAN, was received at TAN via shipments of radioactive 

materials (e.g., irradiated fuels, contaminated equipment), was present as non-dispersible radioactive 

sources, and may have been present as liquid laboratory standards. With the exception of nondispersible radioactive sources and liquid laboratory standards, the known plutonium at TAN was 

always present with mixed fission products. 

5.2.3.5 Neptunium

Neptunium has been present at TAN since radiological operations started there in 1955. Neptunium 

was produced by the various reactors at TAN, and was received at TAN via shipments of radioactive 

materials (e.g., irradiated fuels, contaminated equipment). With the exception of a single shipment of 

Np-237 to TAN in February 1964 (TAN Shipments, 1963-64), the known neptunium at TAN was 

always present with mixed fission products. Very little is known about the Np-237 that was shipped 

in February 1964 other than it was shipped from MTR in solid form and only weighed 8 oz (TAN 

Shipments, 1963-64).

5.2.3.6 Uranium

Uranium has been present at TAN since radiological operations started there in 1955. At TAN, 

uranium was used by the various TAN reactors as fuel, was received at TAN via shipments of 

radioactive materials (e.g., non-irradiated fuels, irradiated fuels, contaminated equipment), may have 

been present as non-dispersible radioactive sources, and may have been present as liquid laboratory 

standards. With the exception of non-dispersible radioactive sources, liquid laboratory standards, 

uranium contamination in the TAN-607 vaults, and potentially, the contamination found in TAN-615, 

the known uranium at TAN was always present with mixed fission products. 

NOTE: During research for the SEC-00219 evaluation report, it was discovered that uranium 

contamination without mixed fission products was present in the Fuel Storage Vaults of the 

TAN-607 and TAN-615 buildings. Further investigation is needed to determine the source of 

the contaminants and the beginning and end dates for the presence of the uranium 

contamination. 

5.2.3.7 Other Actinides

Other actinides (i.e., excluding plutonium, neptunium, and uranium) have been present at TAN since 

radiological operations started there in 1955. At the TAN, other actinides were produced by the 

various reactors at TAN, and were received at TAN via shipments of radioactive materials (e.g., 

irradiated fuels, contaminated equipment, etc.). Based on the available information, the other 

actinides were never present at TAN without mixed fission products being present. 

5.2.3.8 Other Radionuclides

The “other radionuclides” category includes any other radionuclides that were not addressed in the 

radionuclide categories listed above. This category includes fission and activation products that had 
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been chemically separated from the other mixed fission and activation products. With the exception 

of the radionuclides that are specifically addressed below, the majority of the “other radionuclides” at 

TAN were likely encapsulated experiments or non-dispersible radioactive sources.

Tritium (H-3) is produced by the neutron bombardment of a number of elements, and was likely 

present as a contaminant in some TAN locations. Sodium-24 (Na-24) is produced by the neutron 

bombardment of stable sodium, and was likely present as a contaminant in any of the TAN’s 

sodium-cooled reactors or as part of the sodium contamination of any sodium-cooled reactor parts that 

came from TAN or other areas (onsite and offsite). Carbon-14 (C-14) can be produced from a number 

of neutron interactions, and was likely present as a contaminant in carbon, nitrogen, and/or oxygenbearing materials that were irradiated in nuclear reactors at TAN or other areas (onsite and offsite). 

5.2.4 Internal Radiological Exposure Sources from Misc. Reactor Areas Operations

The miscellaneous reactor areas include the Auxiliary Reactor Area (ARA), Organic Moderated 

Reactor Experiment (OMRE), and the Special Power Reactor Experiment Tests. While each project’s 

purpose was unique, the radiological sources for internal exposure to personnel had much in common, 

with the notable exception being ARA-I. ARA-I did not house a reactor, but did have hot cells in 

support of the Army’s boiling-water and gas-cooled reactor programs, and later supported Test 

Reactor Area research activities. Radionuclides with the widest application throughout the 

miscellaneous reactor areas during the period under evaluation are listed below. They are discussed in 

the following subsections and in Section 7.2. 

• Fission and Activation Products 

• Exotic radionuclides (e.g., C-14, Ca-45, P-32, S-35, Ba-140/La-140, and others produced by 

reactor neutron irradiation)

• Radioactive Noble Gases

• Iodine 

• Uranium 

5.2.4.1 Fission and Activation Products

Fission and activation product exposure potential has existed at INL since the site’s inception because 

of its mission as the “National Reactor Testing Station.” Radionuclides created from nuclear fission 

and activation of equipment materials created a variety of beta- and beta-gamma emitting 

radionuclides. For the miscellaneous reactor areas, fission and activation products were the most 

significant sources of internal radiation dose. This fact is reflected in the radiobioassay, radiological 

survey methods, and air sampling programs, which focused on “mixed fission product” monitoring at 

those facilities. 

Fission and activation product exposure occurred as a result of the SL-1 accident. Personnel involved 

in event response and recovery operations were exposed to airborne radioactive material, including 

fission and activation products, such as Sr-90 (SL-1 Sr-90 Dose, 1961).
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5.2.4.2 Exotic Radionuclides

ARA-I: In 1968, separation of Pa-233 from irradiated thorium slugs from MTR was performed. The 

facility was modified for the separations work, including upgrades to the ventilation system and 

addition of isolation boxes. To date, personnel monitoring data have not been found by NIOSH; 

consequently, this Pa-233 work in ARA-I in 1968 is being reserved for further evaluation.

OMRE: Of the many postulated benefits of the use of liquid hydrocarbons as a moderator and coolant 

for a reactor was a reduction in the quantities of induced radioactivity. This proved to be true along 

with other favorable properties, such as low operating pressures resulting in fewer leaks and the rapid 

solidification of the liquid hydrocarbons if they did leak. A study of the radioisotopic composition of 

the OMRE coolants after continuous reactor operation from January 1958 to August 1960 verified this 

benefit. The study measured the radionuclides listed in Table 5-11. The principal gamma-emitting 

isotopes were Mn-56, Fe-59, Mn-54, As-76, and Se-75. The primary beta-emitting isotopes were 

P-32, H-3, S-35, Fe-55, and C-14. Activation of trace impurities is believed to have been the source 

of most of the radionuclides identified, although C-14 and H-3 activity was judged to be from 

activation of the organic coolant (OMRE, 1961). 

Table 5-11: Radionuclides Measured in OMRE Coolant

Radionuclide Half-Life Activity in Coolant (μCi/gm)

Mn-56 2.6 h 3.8E-01

P-32 14.22 d 2.8E-01

H-3 12.2 y 1.8E-01

S-35 87 d 3.8E-02

Fe-55 2.6 y 3.0E-02

C-14 5568 y 2.5E-02

Fe-59 44.3 d 1.7E-02

As-76 26.5 h 9.9E-03

Mn-54 300 d 7.3E-03

Se-75 121 d 7.2E-03

Na-24 14.9 h 1.7E-03

Co-58 71.3 d 1.6E-03

Cu-64 12.8 h 1.6E-03

Ba-140 12.8 d 8.5E-04

Sb-122 2.75 d 8.1E-04

I-131 8 d 8.1E-04

Co-60 5.24 y 7.8E-04

Zn-65 246 d 5.9E-04

Cs-137 30 y 4.5E-04

Zr-95 65 d 3.3E-04

Sr-90 27.7 y 7.7E-05
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5.2.4.3 Radioactive Noble Gases

Radioactive noble gases were produced at the Misc. Reactor Areas but typically in small quantities, as 

shown in the example provided in Table 5-12. Noble gases did not represent a major internal 

exposure concern. ARA and SPERT both used a central control room at distances of approximately 

500 meters and 0.5 miles, respectively, from the reactors they operated. In addition to distance, 

SPERT relied on site security to monitor the SPERT perimeter during testing, as well as

meteorological data for taking radiological measurements downwind from releases (SPERT and 

SNAPTRAN, 1960s; SPERT Dosimetry, 1957-58). 

Table 5-12: Example Air Effluent Inventory for SL-1 (ARA-II)

Time Period Radionuclides Total Activity (Ci)

3rd Quarter 1959 Kr-85, Kr-88, Xe-133, Xe-135, Xe-138 44.13

4th Quarter 1959 Kr-85, Kr-88, Xe-133, Xe-135, Xe-138 73.10

1st Quarter 1960 Kr-85, Kr-88, Xe-133, Xe-135, Xe-138 41.5

2nd Quarter 1960 Kr-85, Kr-88, Xe-133, Xe-135, Xe-138 59.1

Source: SL-1 Safety, 3Q-1959; SL-1 Safety, 4Q-1959; SL-1 Safety, 1Q-1960; SL-1 Safety, 2Q-1960

After the first destructive test at SPERT I, only the daughters of noble gases were detected by gamma 

spectroscopy of downwind air samples. The following radionuclides were detected: Sr-91, Y-91, 

Sr-92, Y-92, and Ba-139 SPERT-I, 1965). 

As discussed in the introduction to Section 5.2.2, because radioactive noble gases do not represent a 

significant internal exposure concern, NIOSH finds there is no need to assess the internal doses 

attributable to radioactive noble gases. 

5.2.4.4 Iodine

Isotopes of iodine are produced in high yield during fission; their short half-lives permit equilibrium 

concentrations to be built up rapidly. Considerable inventories of these isotopes are present after only 

a short duration of reactor operation. Because radioiodines are gases, they are subject to release if 

fuel-element-cladding integrity is breached. Consequently, radioiodines were not typically a concern 

for internal radiological exposures for the miscellaneous reactor areas. However, radioactive isotopes 

of iodine, primarily I-131, were detected in gross amounts after the SL-1 accident (Nelson, 1961). 

Three-hundred-sixty urine samples were collected from 180 workers and analyzed for I-131 by 

gross-gamma counting post-event. Fourteen workers were placed on follow-up urine sampling 

programs to determine urinary excretion (SL-1 Report, 1961). In addition, thyroid counts were 

performed at the CPP Dispensary on all workers associated with the SL-1 event (SL-1 Chronology, 

1961).
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5.2.4.5 Uranium

ARA: Uranium exposure potential at ARA was primarily centered at the ARA-I hot cells where 

irradiated fuel was examined; however, due to remote handling, this potential was considered small. 

After shutdown and a 24-hour period for radiation levels to decline, the ARA reactor fuel elements 

would be transferred to a handling cask for transport to the ARA hot-cell facilities for examination. 

The steam explosion and subsequent core meltdown at SL-1 on January 3, 1961 resulted in the release 

of a multitude of radionuclides, including uranium. SL-1 used a highly-enriched uranium fuel core 

that resulted in dispersion at the accident site. Analysis of samples taken from SL-1 post-incident and 

analyzed at the Central Facilities radiochemistry laboratory showed the uranium to be composed of 

the following radionuclides by mass: 1% U-234, 91% U-235, 2% U-236, and 6% U-238 (SL-1 Report, 

1961). 

SPERT: Only the planned destructive tests at SPERT I between November 1962 and April 1964 

showed any damage to the uranium cores used. The first SPERT I destructive tests used highlyenriched uranium with an aluminum cladding, while the second and third destructive tests used a 

low-enrichment UO2 pin core. The first test resulted in damage to the core but without measureable 

uranium released. The second and third tests showed little damage to the uranium fuel pins 

(SPERT-I, 1965). 

5.2.5 Internal Radiological Exposure Sources from CFA Operations

Radionuclides with the widest application throughout the Central Facilities Area during the period 

under evaluation are listed below. They are discussed in the following subsections and in Section 7.2. 

• Fission and Activation Products 

• Uranium 

Any other potential exposures from exotics or other radionuclides would be dominated by the mixed 

fission products and uranium exposures at the CFA Laundry Facility. 
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5.2.5.1 Fission and Activation Products

As fission and activation products were the primary contaminants at INL, protective clothing would, 

as expected, become contaminated during use. Contaminated laundry generated at INL was sent to 

the CF-669 Laundry Facility in the Central Facilities Area for cleaning. Laundry personnel were 

cognizant of the mixed fission product hazard as Safety Work Permits (SWPs), protective clothing, 

and film badges were required precautions. Paperwork, including contamination levels, was required 

to be completed by all contaminated-laundry generators. Laundry was required to be sorted by type 

(coveralls, shoe covers, etc.) and contamination level (0-10 mR/hr, 10-100 mR/hr, and >100 mR/hr)

(Savignac, 1961, PDF p. 13). There was a dirty-laundry reception area equipped with a radiation 

detector above the entry door and a CAM in the laundry area itself.

5.2.5.2 Uranium

Central Facilities had a fully-equipped machine shop (CF-640) that was used for work that other plant 

areas could not handle. The materials brought to the machine shop had to be of low radiation and 

contamination levels (Stroschein, 1967). An example of a uranium source was un-irradiated SPERT 

fuel that was sent to CF-640 for work that required minor drilling (Shipment Records, 1960s). 

5.2.6 Internal Radiological Exposure Sources from Burial Ground Operations

Radionuclides with the widest application for the Burial Ground during the period under evaluation 

are listed below. They are discussed in the following subsections and in Section 7.2. 

• Fission and Activation Products 

• Plutonium 

Any other potential exposures from exotics or other radionuclides would be dominated by the mixed 

fission product exposures at the Burial Ground. In late 1969, some experimental equipment that was 

accidentally buried in Pit 1 was retrieved; this marked the first time in the history of the Burial 

Ground that buried waste was deliberately exhumed. Further NIOSH evaluation is required of the 

personnel monitoring data for waste-retrieval activities, necessitating reservation of the Burial Ground 

from 1969 forward. 
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5.2.6.1 Fission and Activation Products 

INL-generated waste was virtually all mixed-fission-product waste and was buried in trenches as 

low-level waste (LLW). The volume and curie content of the LLW are presented in Table 5-13. Most 

routine wastes were packaged in cardboard boxes, sealed with tape, and placed in designated 

Dempster Dumpster containers for disposal. By 1957, a routine schedule had been established to pick 

up INL-generated solid waste (RWMC History, 1985). 

 

Table 5-13: Solid Radioactive Waste at INL Burial Ground

Year

Buried LLW Buried TRU Stored TRU

Volume (m3) Ci Volume (m3) Ci Volume (m3) Ci

1952-1960 18,475 60,920 10,545 11,300 NA NA

1961 6,091 155,650 2,439 3,650 NA NA

1962 5,730 115,320 2,439 3,780 NA NA

1963 5,445 251,480 2,755 10,520 NA NA

1964 3,132 146,330 3,357 12,270 NA NA

1965 4,076 685,890 3,764 17,010 NA NA

1966 4,634 859,010 3,454 65,290 NA NA

1967 3,820 836,130 4,859 41,670 NA NA

1968 3,947 268,810 5,826 32,690 NA NA

1969 4,740 935,520 9,791 35,480 NA NA

1970 4,151 482,640 8,429 15,460 1,420 4,225

Source: RWMC History, 1985

Note: All radioactivity values are at time of disposal or storage. 

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5.2.6.2 Plutonium

The INL Burial Ground began receiving shipments of Rocky Flats waste in 1954 and continued to 

receive routine shipments past the end of the period under evaluation for this report. Table 5-14 

presents estimated quantities of the radionuclides received from Rocky Flats between 1954 and 1970. 

Table 5-14: Estimated Quantities of Radioactive Materials From Rocky Flats

Year U-238

(kg)

U-235

(g)

Pu-238

(g)

Pu-239

(g)

Pu-240

(g)

Pu-241

(g)

Pu-242

(g)

Am-241

(g)

1954 733 180 0.02 239.09 14.94 0.90 0.04 14

1955 979 870 0.15 1,529.26 95.54 5.79 0.26 26

1956 1,174 1,108 0.22 2,266.23 141.59 8.58 0.38 28

1957 2,147 1,616 0.30 3,184.16 198.94 12.05 0.54 43

1958 4,209 5,009 0.41 4,320.55 269.93 16.36 0.74 61

1959 3,753 11,319 0.46 4,800.61 299.93 18.18 0.82 62

1960 4,123 16,481 0.34 3,553.84 222.06 17.15 0.61 46

1961 4,311 13,921 0.53 5,538.71 346.09 26.73 0.95 70

1962 4,674 4,493 0.50 5,272.52 323.14 24.95 0.88 180

1963 1,672 6,269 1.53 15,944.51 996.30 76.94 2.72 204

1964 1,339 7,888 1.79 18,586.94 1,161.41 89.69 3.17 238

1965 4,269 4,066 2.43 25,340.54 1,583.41 122.28 4.33 414

1966 53,452 1,688 8.73 90,820.55 5,674.95 438.25 15.51 2,815

1967a 53,176 1,889 5.4 56,231.04 3,513.61 271.34 9.60 2,129

1968 33,373 1,210 4.18 43,543.44 2,720.83 210.11 7.44 1,778

1969 22,721 36,259 3.25 38,046.37 2,376.93 149.34 6.49 4,875

1970 7,084 14,196 2.58 24,178.81 1,517.63 105.82 5.16 1,834

Totals 203,194 128,467 32.82 343,296.17 21,457.23 1,594.46 59.64 14,817 a

 Includes 56 g U-233 (RFP Waste, 2005) 
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Waste from Rocky Flats primarily came in the form of sealed 55-gallon drums, but 30-gallon drums 

were also used. In addition, crates and wooden boxes were used for oversized waste shipments 

(RWMC History, 1979). Rocky Flats waste was buried in large quantities at one time, as recorded on 

monthly off-site radioactive waste disposal reports (Waste Shipments, Feb1963; Waste Shipments, 

Mar1963; Waste Shipments, Apr1963; Waste Shipments, May1963; Waste Shipments, Jun1963; 

Waste Shipments, Jul1963; Waste Shipments, Aug1963; Waste Shipments, Nov1963). The Central 

Facilities health physicist was required to be present for all waste emplacements and performed 

contamination surveys before and after the job. 

In 1962, a complete survey of the Burial Ground was performed after the “Chinook” flood event 

caused the movement and disarrangement of some waste. Particular attention was paid to open Pits 

#2 and #3 and open Trenches #25 and #26, which had filled with water during the flood and 

overflowed. The only alpha contamination found was 5,000 dpm on the sides of Pit #2 and 50,000 

dpm at the bottom of Pit #2. No alpha contamination was found outside the pit. It should be noted 

that the isotopic composition of the alpha contamination was not mentioned, but the initial burial pits 

were excavated for oversized waste from Rocky Flats (HP Monthly Reports, 1962). 

5.3 External Radiological Exposure Sources from INL Operations

To provide consistency of radiation safety programs at the INL amid a large variety of facilities and 

constantly-changing contractors, the AEC established a Health and Safety (H&S) Laboratory at INL 

to provide technical support for internal and external dosimetry programs. The name of this 

organization was changed to Health Services Laboratory (HSL), then to the Health and Safety 

Division (HSD), then to the Idaho Center for Radiological and Environmental Sciences, and most 

recently, to the Radiological and Environmental Sciences Laboratory (RESL) (ORAUT-TKBS-0007-

6).

By policy, INL monitored by dosimeter all personnel who were expected to receive any radiation dose 

or whose work was centered at the site (Cipperley, 1968; CPP HP Manual, 1952; Cipperley, 1958). 

Dosimeters were usually stored at the operational area entrance security gates. 

Table 5-15 shows a historical summary of INL beta-gamma dosimetry systems and their 

corresponding Limits of Detection (LODs). 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table 5-15: INL Beta-Gamma Dosimetry Systems History and LOD Summary 

Period of Usea Dosimeter Exchange 

Frequency 

LODb (mrem) 

Beta Gamma 

August 1951– 

March 1958 

INL initial film, 552 DuPont film Weekly (n=52) 30 30 

Monthly (n=12) 

Reactor areas, DuPont 558 film Weekly (n=52) 30 10 

March 1958– 

December 1966 

INL multielement DuPont 508 film Weekly (n=52) 30 10 

Biweekly (n=26) 

Monthly (n=12) 

December 1966– 

February 1974 

INL multielement DuPont 508 film Weekly (n=52) 30 10 

Biweekly (n=26) 

Monthly (n=12) 

INL LiF TLD Quarterly (n=4) 15 15 

Semi-ann (n=2) 

Annual (n=1) 

February 1974– 

May 1975c 

INL Atlas TLD LiF in Teflon Monthly (n=12) 30 30 

Quarterly (n=4) 

Semi-ann(n=2) 

Annual (n=1) 

December 1974-

December 1985c 

INL Harshaw two-chip TLD Monthly (n=12) 15 15 

Quarterly (n=4) 

Annual (n=1) 

January 1986–2006 INL Panasonic four-chip TLD Monthly (n=12) 15e 15d 

Quarterly (n=4) 

Monthly (n=12) 30e 10d 

Quarterly (n=4) 

Source: This table derived from Table 6-15 in ORAUT-TKBS-0007-6. 

a.		 For many years, INL workers had a dosimeter assigned to each operating area where they worked, 

or they were issued visitor dosimetry. All-area dosimetry was issued beginning in January 2000. 

b.		 LODs are based on: Cipperley, 1958; Cipperley, 1968; Cusimano, 1963; Kalbeitzer, 1983; Gesell, 

1986; Gesell, 1992; Perry, 1993; Ruhter , 2002. 

c.		 ICPP began using the Harshaw TLD in December 1974, the prime contractor began in February 

1975, and ANL-W began in May 1975. 

d.		 The LOD was 15 mrem from January 1, 1986 to July 7 1986; 10 mrem from July 7, 1986 to 

September 1989; and 15 mrem until 1993, when it returned to 10 mrem. 

e.		 The LOD was 15 mrem from January 1, 1986 to July 7 1986 and 30 mrem after that. 

Table 5-16 shows a historical summary of INL neutron dosimetry systems and their corresponding 

Limits of Detection (LODs). 

Table 5-17 lists the principal gamma-emitting fission products that could be expected in the primary 

feed and waste solutions. NOTE: The list begins in the left half of the table and continues in the right 

half of the table. 

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Table 5-16: INL Neutron Dosimetry Systems History and LOD Summary 

Period of Use Dosimeter Exchange 

Frequency 

LOD 

(mrem) 

1951–1958 NTA Film (Kodak Type A Film) Weekly (n=52) 14a 

1959–September 1976 NTA Film (Kodak Type A Film) Weekly (n=52) 20 

Biweekly (n=26) 

Monthly (n=12) 

January 1986–2006 INL Panasonic four-chip TLD Monthly (n=12) 15b 

Quarterly (n=4) 

Monthly (n=12) 

Source: This table derived from Table 6-16 in ORAUT-TKBS-0007-6. 

a. Cipperley, 1958, PDF p. 20 

b. Gesell, 1996, PDF p. 19 

Table 5-17: Mixed Fission Product Gamma and Beta Emitters 

Nuclide Half-Life 

(y) 

Photon 

Energy 

(MeV) 

Max. Beta 

Energy 

(MeV) 

Nuclide HalfLife 

(y) 

Photon 

Energy 

(MeV) 

Max. Beta 

Energy 

(MeV) 

Selenium-79 3.27 x 105 - 0.149 Samarium-151 96.6 - 0.076 

Krypton-85 10.76 - 0.687 

Europium-155 4.76 

0.105 0.128 

Strontium-90 28.9 - 0.546 - 0.141 

Zirconium-93 1.53 x 106 - 0.062 - 0.159 

Niobium-95 (a) 35 d 0.765 0.160 - 0.186 

1.121 - - 0.246 

Technetium-99 2.11 x 105 - 0.294 

Terbium-160 72.3 

0.197 0.434 

Rhodium-106 (b) 368.2 d 

0.512 1.979 0.216 0.475 

0.622 2.407 0.299 0.546 

1.051 3.029 0.392 0.569 

- 3.541 0.765 0.784 

Palladium-107 6.5 x 106 - 0.033 0.879 0.867 

Cadmium-113m 14.1 - 0.586 0.962 1.747 

Tin-126 2.3 x 105 - 0.250 0.966 -

Iodine-129 1.57 x 107 - 0.152 1.115 -

Cesium-135 2.3 x 106 - 0.205 1.178 -

Cesium-137 30.23 - 0.511 1.200 -

- 1.173 1.272 -

Barium-137m (c) 2.552 min 0.662 - 1.312 -

Praseodymium144 (d) 17.28 min 

0.696 0.810 

- 2.300 ---

- 2.996 

Source: CPP Fact Sheet, undated 

Notes: 

(a) Decay product from zirconium-95 

(b) Decay product from ruthenium-106 

(c) Decay product from cesium-137 

(d) Decay product from cerium-144 

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5.3.1 External Radiological Exposure Sources from Test Reactor Area Operations

5.3.1.1 Photon

Photon exposure resulted from mixed fission and activation products and isotopes produced by 

reactors and from separations of Pa-233 performed in the MTR alpha cave. Exposure rates varied by 

dates, but generally, exposure potential increased with time due to increased production and the 

increasing number of reactors, in 1957 and in 1967. The radiation fields at TRA, with a few 

exceptions, were generated primarily by mixed fission and activation products. Therefore, most of the 

photon dose has been from photons with energy greater than 250 keV (ORAUT-TKBS-0007-6, PDF 

p. 26). 

5.3.1.2 Beta

Beta exposure resulted from mixed fission and activation products and isotopes produced by reactors. 

Exposure rates varied by dates, but generally, exposure potential increased with time due to increased 

production and the increasing number of reactors, in 1957 and in 1967. The radiation fields at TRA, 

with a few exceptions, were generated primarily by mixed fission and activation products. Beta dose 

resulted primarily from exposure to electrons with greater than 15 keV (ORAUT-TKBS-0007-6, PDF 

p. 28). 

5.3.1.3 Neutron

Test Reactor Area workers may have been exposed to neutrons from the existing nuclear reactors and 

from neutron sources. Shipping transactions indicate the likely receipt of polonium-beryllium and 

radium-beryllium sources though NIOSH has not located documentation on number and strengths of 

such sources. Table 5-1 lists all TRA reactors in existence from 1952 through 1970. There was also 

potential for neutron exposure from the Fast Chopper and Crystal Spectrometer.

5.3.2 External Radiological Exposure Sources from CPP Operations

5.3.2.1 Photon

Photon exposure resulted from the mixed fission and activation products from the reactor fuels 

processed at CPP. The workplace photon energy spectra at CPP would have been dominated by 

high-energy photons from mixed fission and activation products. Pa-234m is a decay product in the 

U-238 decay chain and emits a 2.29 MeV beta particle. A significant quantity of photons resulting 

from Bremsstrahlung radiation are produced and contribute photons of intermediate energy (30-250 

keV). Although enriched uranium has significantly less in-growth of Pa-234m, U-235 and its decay 

products emit a 185.7 keV photon 57% of time and a 143.8 keV photon 11% of the time. These 

photons dominate the measured photon energy spectra in the product streams. 

The majority of radioactivity in process raffinate solutions and calcined wastes were from the long 

and medium half-life fission products. The most significant gamma activity would have been from 

cesium-137 decaying to barium-137m and its subsequent gamma emission, as well as the decays of 

europium-154, ruthenium-106, and cerium-144. To a lesser degree, gamma activity would also result 

from the decay of rhodium-106 and praseodymium-144. 
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Because fuels used in the RaLa process were cooled for only two days rather than the 90-120 days 

used for other process streams, there would be a large population of short-lived fission products 

present in the fuel and wastes. 

In addition to the fission product gamma rays, gamma rays from the activation of fuel cladding and 

structural materials, such as Mn-54, Co-58 and Co-60, would also be present, but would vary 

depending on the type of fuel being processed at the time. 

Each of the process cells was constructed with five-foot-thick concrete walls to shield operators from 

the radiation fields. The penetrations of the walls contained bends, thereby preventing radiation 

leakage from the inside the cell. However, radioactive process fluids may have gotten into exposed 

instrument leads and parts of the pulse pump systems that normally would be clean. The introduction 

of these highly-radioactive fluids into lines in the Operating Corridor and the Access Corridor could 

give rise to high radiation fields (Stroschein, 1967).

Table 5-18 lists the mixed fission products resulting from the RaLa process. NOTE: The list begins in 

the left half of the table and continues in the right half of the table.

Legend For Table 5-18 (next page): 

(a) Emissions with yields of less than 1% or energies of less than 100 keV are not reported

(b) Te-131 emits 93 different photons. Only those with yields greater than 10% are reported here.

(c) Promethium-151 emits 76 different photons. Only those with yields greater than 10% are reported.
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Table 5-18: Mixed Fission Products from the RaLa Process 

Nuclide Half-Life 

(d) 

Photon 

Energy 

(MeV) (a) 

Max. Beta 

Energy (MeV)(a) Nuclide Half-Life (d) 

Photon 

Energy 

(MeV) (a) 

Max. Beta 

Energy 

(MeV)(a) 

Sr-89 50.55 0.909 1.491 

La-140 1.68 

0.329 1.239 

Y-91 58.51 1.205 1.543 0.433 1.244 

Zr-95 64.02 0.724 0.366 0.487 1.279 

0.757 0.399 0.752 1.296 

Nb-95 35.00 0.765 0.160 0.816 1.348 

1.121 0.868 1.412 

Ru-103 39.35 

0.497 0.113 0.920 1.677 

0.610 0.226 0.925 2.164 

0.723 1.597 -

Rh-105 1.47 

0.306 0.248 2.522 -

0.319 0.261 Ce-141 32.5 0.145 0.435 

0.567 - 0.580 

Ag-111 7.46 

0.245 0.686 

Ce-143 1.38 

0.232 0.517 

0.342 0.723 0.293 0.733 

- 1.028 0.351 1.104 

Cd-115 2.23 

0.261 0.184 0.490 1.398 

0.336 0.197 0.665 -

0.492 0.267 0.722 -

0.528 0.394 Pr-143 13.56 - 0.935 

Sn-125 9.64 

0.332 0.348 

Nd-147 10.98 

0.319 0.210 

0.270 0.367 0.440 0.365 

0.823 0.460 0.531 0.805 

0.916 1.261 Pm-149 2.21 0.286 0.785 

1.067 2.350 - 1.071 

1.089 -

Pm-151 1.18 

0.340 (c) 0.235 

2.001 - - 0.310 

Te-129m 33.6 

0.275 0.908 - 0.365 

0.031 1.604 - 0.414 

0.696 - 0.447 

Te-131m 1.25 

0.774 0.317 - 0.742 

0.794 0.420 - 0.792 

0.852 0.430 - 0.843 

1.126 (b) 0.451 - 0.864 

- 0.507 - 0.881 

- 0.532 - 0.979 

- 0.544 - 1.020 

- 0.785 - 1.118 

- 2.431 - 1.188 

Te-132 2.26 

0.112 0.215 

Sm-153 1.95 

0.103 0.632 

0.116 - - 0.702 

0.228 - - 0.805 

I-131 8.04 

0.284 0.248 

Eu-156 15.19 

0.646 0.183 

0.364 0.334 0.723 0.248 

0.637 0.606 0.812 0.266 

0.723 - 1.065 0.426 

Xe-133 5.25 - 0.346 1.153 0.487 

Ba-140 12.79 

0.163 0.454 1.154 1.087 

0.305 0.567 1.231 1.211 

0.424 0.872 1.242 1.285 

0.438 0.991 - 1.404 

0.537 1.005 - 2.453 

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5.3.2.2 Beta

Beta exposure resulted from the mixed fission and activation products from the reactor fuels processed 

at CPP. Most fission products are initially rich in neutrons and undergo beta decay. The beta 

emissions for regular process fuels and RaLa fuels are presented in Table 5-17 and Table 5-18, 

respectively. 

In general, the thick shielding afforded by the cell walls would attenuate any beta particles emitted. 

However, at the end of a production run, it was usually necessary to perform maintenance work on the 

process piping, vessels, and instrumentation within the cells (Stroschein, 1967). Cell entry would 

have required the filing of a special work permit (CPP HP Manual, 1952). Before anyone attempted 

to enter the cells, it was necessary to decontaminate the piping and vessels internally. This was done 

by remote control and would result in reduction to acceptable work levels of radiation from the 

process piping (Stroschein, 1967). For a cell entry, an acceptable work level would be a dose-based 

limit of 300 mrem/week.

If there was leakage of radioactive solutions from any part of the process system, the cell walls, the 

floor, and external areas of the piping and vessels may have become highly contaminated. A spray 

system was built into each cell to wash down the process system exteriors and cell surfaces. This was 

designed to reduce the radiation to a level that permitted removal of the top hatch and subsequent 

surveys by Health Physics personnel. 

Despite the above mitigating actions, hot spots may have persisted either inside or outside of the 

process system. It was, therefore, often necessary to provide temporary shielding to allow work to be 

done in the cells. Even so, the allowable work time may have been limited for certain tasks, and the 

health physicist was required to maintain tight control over this aspect of the work (Stroschein, 1967).

5.3.2.3 Neutron

Neutron exposure could have resulted from the separations processes and product lines, which 

handled quantities of TRU isotopes and the associated spontaneous fission and α/n neutrons produced. 

Uranium present in a neutron flux can capture neutrons, producing transuranic nuclides. Many of 

these nuclides produce neutrons through spontaneous fission. Alternately, these nuclides may decay 

by alpha emission which, if this decay occurs in the presence of a low-atomic-number material, may 

lead to the production of neutrons via the (α,n) reaction. 

5.3.3 External Radiological Exposure Sources from TAN Operations

Radiological operations have taken place at TAN since 1955. TAN was primarily a nuclear reactors 

operations area.

5.3.3.1 Photon

Based on a review of the available INL records, the photon radiation fields at TAN were generated 

primarily by the operation of nuclear reactors, activated materials, and irradiated reactor fuels. Photon 

energies were predominately greater than 30 keV (ORAUT-TKBS-0007-6). Photon radiation fields at 

TAN are known to have ranged from immeasurable levels to over 1E+6 mrem/hr. The higher dose 
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rates in this range were from unshielded test reactors where personnel were typically protected either 

by distance, underground shielded bunkers, or both. 

5.3.3.2 Beta

Based on a review of the available INL records, the non-penetrating radiation fields at TAN were 

generated primarily by activated materials, irradiated reactor fuels, and radioactive contamination 

from those materials. Because there is no indication of separated plutonium at TAN (with the 

exception of some laboratory standards), all non-penetrating radiation at TAN is attributable to 

electron (beta) radiation. Electron energies were predominately greater than 15 keV (ORAUT-TKBS0007-6). Electron radiation fields at TAN are known to have ranged from immeasurable levels to 

over 1E+3 mrem/hr.

5.3.3.3 Neutron

Based on a review of the available INL records, the neutron radiation fields at TAN were generated 

primarily by the operation of nuclear reactors and irradiated reactor fuels. Shipping transactions and 

other records indicate the presence of polonium-beryllium, plutonium-beryllium, and 

radium-beryllium neutron sources. However, NIOSH has not located documentation on the number 

and strengths of those neutron sources. Neutron energies were generally less than 20 MeV (ORAUTTKBS-0007-6). Neutron radiation fields at TAN are known to have ranged from immeasurable levels 

to over 1E+6 mrem/hr. The higher dose rates in this range were from unshielded test reactors where 

personnel were typically protected either by distance, underground shielded bunkers, or both. 

5.3.4 External Radiological Exposure Sources from Misc. Reactor Areas Operations

5.3.4.1 Photon

Photon exposure resulted from photon radiation produced during nuclear fission. In addition, many of 

the fission and activation products resulting from fuel irradiation decay by photon emission. The 

workplace photon energy spectra in the Misc. Reactor Areas would have been dominated by 

high-energy photons from mixed fission and activation products. OMRE was slightly different in that 

Mn-56 contributed approximately 80% of the detected radiation from operations (OMRE, 1961). 

The SPERT and Auxiliary Reactor Area reactors were operated from remote-control rooms. One of 

the designed benefits was the control of direct radiation hazards to workers through the following 

mechanisms:

• Evacuation of personnel during reactor operations 

• Health physics and security control of access 

• Remote area monitors with read-outs in control rooms 

• Time delay to allow decay of short-lived radionuclides before reentry 

Based on a review of available records, the highest external exposure potential from photons in the 

Auxiliary Reactor Area occurred at SPERT-I during removal of the fuel elements from the reactor. 

Under normal operations, fuel elements were transferred using a six-foot handling tool, without 
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shielding, with approximately 10 seconds required for the transfer. External dose rates would range 

up to 20 R/hr at 5 feet (Mills, 1957). “Incident reports” were generated for workers whose external 

exposures exceeded an established control limit during the film badge wear period (SPERT Incident, 

1957; McVey, 1956). 

5.3.4.2 Beta

Reactor operations in the Misc. Reactor Areas did not have pure beta external hazards, but 

non-penetrating exposures would have resulted from the mixed fission and activation products 

produced by the reactors. Whether beta radiation was considered an external hazard depended on the 

maximum energy of the beta emission for a given radionuclide, any shielding used, and the use of 

protective clothing (Mills, 1957a). Beta energies were predominantly greater than 15 keV 

(Measurement Procedures, 1983). 

5.3.4.3 Neutron

Operating reactors would have been the primary source of neutron exposure during the period under 

evaluation. These reactors were experimental facilities for measuring shielding factors, reactor 

kinetics, and neutron irradiation. The SL-1 reactor was caused by the improper withdrawal of one of 

the cadmium control rods, resulting in prompt neutrons creating steam formation in milliseconds. In 

addition, neutron sources such as PoBe, RaBe, and PoB sources were used at INL (H&S Bulletin, 

1958). 

5.3.5 External Radiological Exposure Sources from CFA Operations

5.3.5.1 Photon

Though the Central Facilities Area performed a support role for INL operations, there were sources of 

external photon exposure due the following functions: 

• Receipt and survey of radioactive shipments 

• Contaminated laundry facility 

• Machine shop and vehicle maintenance 

Central Facilities served as the transportation center for INL. All incoming and outgoing radioactive 

shipments by rail or trailer trucks were routed through CFA. Radioactive Shipment Records, which 

included radiation and contamination survey results, were required for all shipments; examples are 

available (McCaslin, 1953; Incoming Shipment Records, 1961). 

The CF-699 Laundry facility routinely handled laundry of various external dose rates and 

contamination levels from mixed fission and activation products found on the contaminated laundry. 

Photon dose rates of contaminated laundry were documented on Radioactive Shipment Records for 

shipments to CF-699 (Sommers, 1959). In addition, Waste Disposal Request and Authorization forms 

generated by CF-699 demonstrate measurable photon dose rates for shipments to the Burial Ground 

for disposal (Belnap, 1963). 
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The CF-640 Machine Shop on occasion would receive low-radiation material for work. Photon 

exposures were also possible at the CF-665 Vehicle Maintenance Shop, where transportation 

equipment was serviced and, when necessary, sent to CPP for decontamination (Stroschein, 1967). 

5.3.5.2 Beta

Beta exposure at CFA would have resulted from mixed fission and activation products and any other 

beta-emitting isotopes produced by reactors. Potential for non-penetrating external exposures would 

have been highest at the CF-699 Laundry facility. Beta energies were predominantly greater than 15 

keV (Measurement Procedures, 1983). 

5.3.6 External Radiological Exposure Sources from Burial Ground Operations

5.3.6.1 Photon

Photon exposure at the Burial Ground would have resulted from radioactive waste generated at INL, 

other waste generators between 1960 and 1963 when INL was designated an interim burial ground by 

the AEC, and the Rocky Flats Plant. Photons from INL-generated waste would have been dominated 

by high-energy photons from mixed fission and activation products. The Burial Ground routinely 

accepted INL waste with dose rates not to exceed 500 mR/hr at one meter from the container (Burial 

Ground, 1962). In addition to routine waste, non-routine waste (defined as dose rates greater than 500 

mR/hr at one meter) that could cause excess personnel exposure was transported in special containers 

and transfer vehicles. These disposals were always carried out under the supervision of a health 

physicist and typically scheduled outside normal operating hours. At least up to 1957, no upper limit 

had been set on the level of radiation that could be handled; items of up to 12,000 R/hr were buried 

(RWMC History, 1985). During its years as an interim burial ground, INL received a great deal of betagamma- contaminated waste from other AEC contractors, university, private industry, and the armed 

forces. 

Shipments of Rocky Flats waste began in 1954. Up until 1963, most of the waste drums from Rocky 

Flats were stacked by hand. Beginning in November 1963 and lasting until late 1969, RFP waste 

containers were no longer stacked, but were dumped into the pits to reduce labor costs and minimize 

personnel photon exposures. Consideration was made for Am-241 in-growth from the Pu-241 

constituent of Rocky Flats waste since Am-241 is an external dose hazard due to the 59.5 keV photon 

emitted by its predominant decay mode (Johnson, 1969). In 1967, wastes generated during U-233 

processing were packaged in lead-lined drums to reduce personnel exposure prior to shipment to INL 

(RFP Waste, 2005). 

5.3.6.2 Beta

Non-penetrating radiation at the Burial Ground would have originated from the mixed fission and 

activation products produced at INL and any beta-emitting radionuclides that were sent to the Burial 

Ground when it was an interim storage facility. However, given the shielding afforded by waste 

containers and waste dumping and burial methods, beta radiation as an external hazard would have 

been minimal compared to photon radiation. Beta energies were predominantly greater than 15 keV

(Measurement Procedures, 1983). 
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5.4 Incidents

INL facilities and activities involved experimental reactor design and development, irradiated fuel 

processing, and low- and high-level radioactive waste treatment and disposal. INL’s monitoring and 

analytical programs were designed to initiate an investigation of any potential internal intake as 

indicated by off-normal workplace indicators such as personnel contamination or positive air 

sampling. NIOSH found extensive personal monitoring data for mixed fission and activation products 

(beta/gamma); therefore, only external exposures of regulatory significance, fire and flood events at 

the Burial Ground, significant alpha-contamination incidents, and criticality incidents are presented 

below.

5.4.1 Incidents at the Test Reactor Area

From the onset of MTR operations through the end of the period under evaluation (1970), incidents 

occurred in the TRA that released airborne fission and activation products. Releases at the reactors 

and reactor canals occurred from opening of the reactor top after reactor shutdown, infrequent breach 

of casks and transfer rabbits (HP Monthly Reports, 1959-1960). Similar releases occurred inside the 

MTR hot cells (Johnson, 1965a; Johnson, 1961; Sommers, 1957a). Infrequent releases of fission 

product materials also occurred in the MTR labs. Releases of fission products resulted in airborne and 

surface beta and gamma contamination. After releases of fission products, TRA Health Physics 

monitored workers with potential for internal intakes by urinalysis or whole-body counting. Starting 

in 1954, TRA Health Physics used gamma-ray spectroscopy to identify unknown contaminants from 

incidents. One example was the identification of Cl-38 from an airborne release detected at both the 

MTR reactor top and the main floor on December 16, 1958 (Keller, 1958). Health Physics performed 

follow-up monitoring of areas affected by incidents until conditions were safe for workers to re-enter. 

NIOSH examined documented reports of releases of fission products from 1957 through 1970. Health 

Physics did post-incident follow-up either by air sampling or worker bioassay for each of those 

documented incidents (HP Monthly Reports, 1959-1960). NIOSH found the bioassay results for 

workers identified in the reports in the data supplied by INL. NIOSH closely examined incident 

records involving the release of alpha-emitting radionuclides. The incidents taking place in the TRA 

are discussed below: 

• Unshielded Reactor Component at the MTR, July 23, 1956: A radioactive reactor component at 

the Materials Testing Reactor was not adequately shielded, resulting in radiation exposures above 

current standards to four employees. While the reactor was shut down for scheduled refueling, the 

water level in the reactor tank was lowered, and the component was moved to facilitate the 

insertion and removal of experiments in the reactor. During this procedure, the radioactive 

component was partially exposed. [Number redacted] employees working on the reactor top 

adjacent to the reactor tank opening and two [roles redacted] received radiation exposures. 

[Number redacted] employees received a 21.5 rem exposure, which exceeded the 15-rem-per-year 

standard in effect in 1956 as well as the current-day 5-rem standard. In addition to [number 

redacted] employees received radiation exposures that exceed the current-day standard. These 

employees received radiation exposures of 6.15 rem, 6.2 rem and 10.6 rem. The other [number 

redacted] employees received radiation exposures below 5 rem. 
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• Alpha Incident, MTR Lab 109, January 1956: Over the period of a week, a group of researchers 

from [organization name redacted] were found by Health Physics to not be practicing good 

contamination control in an area not designated as “clean” (Labs 109 and 110). The researchers 

were found drinking sodas at one point, and later, with messy housekeeping. Contaminated trash 

was found in a trash can in Lab 110. On January 23, the researchers were found to be cutting a 

hot sample at the rabbit canal without wearing protective clothing. Some material at the canal was 

found to be contaminated. Airborne contamination was continuously monitored. Health Physics 

monitored the labs for contamination and contaminated areas were subsequently cleaned. The 

[organization name redacted] researchers were given Health Physics indoctrination (Sommers, 

1956).

• Personal Contamination, August 1958: A worker received an open wound while lathing U-235 in 

the [location redacted] Lab. The wound was determined to be contaminated. The worker was 

assayed using urinalysis. No internal intakes were detected (Sommers, 1959a).

• Lab Contamination, May 1960: A wide area of MTR Metallurgical Lab was contaminated from 

work with U-235, although the source was not identified. Smears of alpha contamination gave a 

maximum of 625 dpm/100 cm2. The contaminated area was roped off until decontaminated. 

Bioassay was collected from personnel working in the area though no intakes were detected (HP 

Monthly Reports, 1959-1960).

• Floor and Equipment Contamination, August 1960: The floor in the MTR Metallurgical Lab was 

contaminated from work with U-235. The floor was contaminated to a maximum of 110 dpm/100 

cm2 while equipment was contaminated to a maximum of 56,000 dpm/100 cm2

. Bioassay was 

collected from personnel working in the area. As a result of that incident, Health Physics required 

anti-contamination clothing when doing similar work in the MTR Metallurgical Lab (HP Monthly 

Reports, 1959-1960).

• Leaking Plutonium Capsule, [location acronym redacted], June 1961: A capsule containing 

plutonium leaked in the [location acronym redacted] contaminating a small area of the facility to 

an average smear of 5,600 dpm/100 cm2. All personnel involved were surveyed and found to be 

free of contamination except one worker. The worker was sent to the CFA dispensary for 

decontamination. Bioassay samples were collected from personnel involved in the incident. The 

results showed no detectable alpha activity (Johnson, 1961).

• Contaminated Capsule Holders, RMF, July 1961: Capsule holders used at RMF became 

contaminated up to 100,000 dpm plutonium. The holders were taken to the Health Physics office 

at TRA. The HP office was subsequently contaminated along with RMF and HP staff, with 

[number of people redacted] [body part redacted] becoming highly contaminated. Contaminated 

areas and staff were easily decontaminated with the exception of [number of people redacted] 

[body part redacted]. That [number of people redacted] [body part redacted] was successfully 

decontaminated after several decontamination iterations over the day. Urine samples were 

collected by all personnel involved and showed no detectable plutonium (Johnson, 1961).
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• Plutonium Contamination, Wide Area of TRA, November 1963: Widespread alpha contamination 

was discovered on the evening shift of November 15, 1963, and subsequently through November 

18, 1963 (Johnson, 1963). This widespread contamination was initially discovered during a 

routine radiation survey of tools brought to the MTR Health Physics office. Contamination from 

the same cause was found in the mock-up area in the MTR service wing, the MTR welding shop, 

and in a major portion of the ETR complex, ETRC, TRA maintenance building, TRA cafeteria, 

and on several walkways and roadways in exclusion areas. Low-level contamination was also 

found on personal items in eight homes and [number and ownership redacted] vehicle (Loop, 

1963a). 

MTR Health Physics found the original cause of the contamination was a contaminated movable 

welding glove box that had not been approved for work with radioactive materials nor surveyed 

after use. The glove box had been used in late-October in the MTR mock-up area (Loop, 1963) to 

weld several thin capsules containing Pu-240 for future irradiation at EBR-1. A drawing of the 

moveable glove box is shown in Figure 5-50. The structure and size of the capsule containers was 

such that they were easily breached during welding. Both maintenance and HP workers observed 

breaches but the glove box was not surveyed for contamination. That glove box was used for 

welding other materials after being contaminated. As those materials were transferred across 

MTR, labs, ETRC and ETR, Pu-240 contamination picked up during welding was transferred 

across these areas. After the contamination was detected, control measures were instituted, which 

included isolation of the area and contamination checks of persons, buses, and homes. Low level 

contamination was found in eight homes, [number redacted] automobile, and three NRTS buses, 

all of which were easily cleaned. Work was barred from affected areas until each was 

decontaminated. The welding glove box was also decontaminated. Workers were given 

additional training in use of safe work permits and in the shipment of radioactive materials. 

Ninety-nine bioassay samples were collected from 78 workers exposed to the contamination. 

Sixty-eight fecal samples were collected from twenty-two workers. Both urine and fecal samples 

were analyzed for plutonium (Horan, 1963). The contamination was documented by INL in 

Findings of Accident Review Committee Alpha Contamination Incident Test Reactor Area (TRA) 

November 15-18, 1963 (Latchum, 1963).
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Figure 5-50: Schematic of Moveable Glove Box (Initial Cause of Pu-240 Contamination) 
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The Atomic Energy Commission performed a separate review of the incident (Loop, 1963) listing 

findings, including one of personnel exposures: 

- Data from urine analyses were positive only on five persons and in all cases were sufficiently 

low to indicate negligible bone dose (plutonium was in the form of an insoluble oxide). 

- Data from positive fecal analyses on eight individuals indicated inhalation to be sufficiently 

low such that maximum dose to the lungs using the established ICRP method and model of 

that era was less than INL’s radiation protection guide.

• Contamination, MTR Alpha Lab, December 1964: Lab 126 of the MTR alpha lab extension was 

contaminated by an unidentified source. The low-level contamination was noted outside glove 

boxes, inside hoods, in a sink, and on one lab hot plate. While the activity was considered 

low-level, Health Physics stressed the need for more stringent contamination control by alpha labs 

staff (Johnson, 1965).

• Contamination, MTR Metallurgical Lab, September 1965: Low-level alpha contamination from 

uranium (about 3 cpm/100cm2

) was discovered in the “cold met lab” at MTR following a spill. 

Furthermore, a [role redacted] retrieved a package from the lab, thus transferring some low-level 

alpha contamination to a [type redacted] vehicle. Contamination was cleaned to below detectable 

level in both the lab and the car. No bioassay was performed given the low-level activity 

(Hawley, 1965).

• Alpha Contamination, MTR, August 1967: Alpha Labs 124, 127, and 128 were contaminated from 

work with americium, curium and plutonium (Phoenix,1967). Smears obtained from surveys of 

contaminated areas ranged as high as 100,000 dpm/100 cm2

. The contamination was subsequently 

tracked by foot to the hallways, other labs in MTR, and to the TRA cafeteria. A glove box in 

Alpha Lab 128 was found to be the original source of the alpha contamination. It was bagged, 

surveyed, and shipped to the Burial Ground. Contamination in the labs, hall, and cafeteria was 

cleaned. Bioassay samples were collected from lab staff exposed to the contamination with a 

potential for intake. 

• Plutonium Air Activity, Lab 128, September 1969: Plutonium air activity was released while 

[number and role redacted] were transferring Pu-238 samples from a glove box to a hood. A 

nearby CAM alarmed with an increase of 1,000 counts per minute. An investigation showed there 

had been little negative pressure in the glove box. [Number and role redacted] were assessed for 

intakes by bioassay; none were found. A corroded exhaust line valve was replaced on the glove 

box (Howe, 1986).

• Americium Alpha Contamination, Lab 128, October 1969: Contamination (identified as Am-241) 

was spread across the lab floor after a boiling chip containing Am-241 exploded. The lab and 

adjacent hallway were sealed off until the areas were decontaminated. Although the [role and item 

redacted] were contaminated, [body part redacted] and [item redacted] were not. Follow-up 

bioassay was negative (Hughes, 2002). 
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5.4.2 Incidents at the Chemical Processing Plant

Since 1949, three events at CPP have caused workers to receive radiation doses in excess of current 

radiation protection standards. 

• Open Transfer Valves, March 20, 1958: Eleven workers at the Idaho Chemical Processing Plant 

received radiation exposures during a routine transfer of radioactive waste material (including 

iodine-131) to permanent storage. The waste materials were being forced under steam pressure 

from one process cell through another process cell to a permanent storage tank. At the time, the 

second cell had been out of use for about eight months. Although the workers had checked the 

cell's process valves and thought them to be closed, the valves, in fact, were stuck partially open 

but gave the appearance of being closed. When the steam pressure built up, radioactive vapor was 

vented through the partially open valves into the work area. Although 11 workers were exposed to 

the radioactive vapor, only seven workers received radiation doses to their thyroids in excess of 

the 30 rem/year thyroid dose standard in effect in 1958. The exposures to the seven workers 

ranged from 30 rem to 210 rem. However, only three of the seven workers would have received 

doses exceeding the current thyroid dose standard of 50 rem/year (Rich, 1958).

• Criticality Accident, October 16, 1959: The accidental transfer of a uranyl nitrate solution to an 

unsafe storage tank resulted in a criticality event at CPP that exposed [number redacted]

individuals to radiation. The liquid solution, which contained enriched uranium, was accidentally 

transferred from a geometrically safe storage tank to an unsafe waste collection tank through a line 

normally used to transfer decontaminating solutions to waste storage. The transfer occurred as a 

result of an inadvertent siphoning action. The siphoning action drew about 34 kilograms of 

enriched uranium (in solution) to the unsafe storage tank, placing the uranium in a critical 

configuration (Ginkel, 1959). The incident spread radioactivity throughout the building and into 

operating areas via vent lines and drain connections, triggering radiation alarms and prompting the 

evacuation of the building. Twenty-one individuals evacuated the process building and the 

surrounding area of high radiation. No indium foils were activated during this incident, indicating 

that any neutron emissions were contained by the shielding (Hayden, 1959). All individuals 

involved in the incident were monitored and, although seven individuals received external 

radiation exposures, none of the individuals received a whole body exposure that exceeded the 15 

rem/year standard in effect in 1959. However, [number redacted] individuals received whole body 

doses of 8 rem and 6 rem, which exceed the current 5-rem whole-body exposure standard (Ginkel, 

1960).

• Maintenance Incident, March 19-24, 1973: The Waste Calcination Facility was in a shutdown 

status for maintenance work on process equipment during the first part of 1973. Equipment 

failure allowing process-solution spillage had resulted in considerable contamination on the cell 

walls, floors, and equipment. Extensive effort had been expended in decontamination over a 

nine-month period, but not all of the radioactivity could be removed. Fission product 

ruthenium-106, resistant to decontamination methods, constituted the major radioactivity 

remaining in the contamination. High radiation doses to a worker in the WCF led to the initiation 

of a joint ACC/AEC-HSL study of problems in dosimetry and dose control at the CPP. Dosimetry 

results from film badges, TLDs, and direct- reading dosimeters were compared. Personnel 

exposure values were also compared with values from controlled experiments. The data showed a 
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serious inconsistency in the relationship between film badges, direct-reading dosimeters, and TLD 

readings. 

A [role redacted] making repairs on the [location redacted] facilities at CPP received high 

radiation exposures. The [role redacted] was on loan to the plant to do repair work on one of the 

[location redacted] contaminated by ruthenium-106. Working in direct contact with the cell, the 

worker was exposed to high levels of high-beta energy fields. Routine film badge checks 

indicated that the worker, who was normally assigned to non-radiation areas, had received high 

exposures. The problem arose, in part, because the radiation monitoring devices (pocket dosimeter 

and film badge) used by the worker were calibrated for radiation fields significantly different from 

the fields encountered. The [role redacted] received a whole body radiation dose of 5.2 rem, which 

was less than the 12 rem/year exposure standard in effect in 1973, but is greater than the current 5-

rem standard (Anderson, 1974, PDF p. 6). 

Periodically, there have been incidents such as spills of transuranic isotopes and excursions within 

shielded cells. Calculated and measured personnel exposures from bioassay results were below 

permissible levels, but each incident re-emphasized to Health Physics and technical personnel the 

importance of critically-safe geometries and the potential for transuranic exposure at the CPP. 

Selected examples of the incidents are listed below.

• Release of Contaminated Air in Z-Cell, January 24, 1958: Contaminated air was released in Z-Cell 

that resulted in exposures to [number redacted] personnel. The contamination was caused by 

excessive air flow in Z-Cell equipment that caused uranium-contaminated fumes to be expelled 

from the cell floor sump jet line. As soon as the vapors were noticed, air samples measuring 2 x 

10-6 μCi/mL were taken near the line and respirators were worn by all personnel entering the cell. 

Urine samples were taken from the [number redacted] men and the [role redacted], and nasal 

swabs were collected from the [number redacted] men involved. No contamination was detected 

on the nasal swabs and the highest urine concentration was 14.8 x 10-5 grams U/L (Sommers, 

1957a). 

• Plutonium Contamination Encountered, July 1959: During the modification of the waste amsco 

storage and transfer system in the access corridor, plutonium-contaminated solution was 

encountered. Although the actual Pu concentration was lower in this case than some average plant 

streams, the Pu was present as a major fraction of the total activity (approximately 50%). At the 

time, the seven personnel involved in this incident were equipped with protective clothing 

(including respirators); hence, no significant internal exposure was incurred. This was confirmed 

by the results of 24-hour urine specimens submitted by each of the seven individuals and analyzed 

for Pu; the results were negligible (Progress Reports, 1959, PDF p. 6). The areas, which were Pucontaminated (i.e., access corridor and solvent burner), were isolated and tagged until 

decontamination was accomplished. Incident investigation suggested that the most probable 

source of the Pu concentration was the interface of the CPM first-cycle extraction columns, which 

had undergone some experimental flow-sheet changes during the previous processing run. No

significant alpha-fallout activity was detected within the plant area, except for the immediate 

solvent burner area (Hayden, 1959a).
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• Plutonium Activity from Burning Solvent, August 1959: An attempt was made during the month 

to burn the Pu-contaminated solvent from first-cycle extraction columns accumulated during the 

past plant run. After 12 hours of burning, Health Physics requested that this operation be 

discontinued because activity was approaching an unsafe limit in the burning area. The solvent 

was later decontaminated and burning was continued without further incident (Hayden, 1959b).

• Criticality Incident in H-Cell, January 1961: A build-up of air pressure in the H-Cell 

decontamination lines lifted a concentrated solution of enriched uranyl nitrate from a 

geometrically-safe section of a first-cycle product evaporator to a critically-unsafe section of the 

evaporator. Of the 251 individuals present in the CPP area at the time of the incident, none 

received a significant radiation exposure. The highest exposure, as determined from film badge 

readings, did not exceed 55 mrem of penetrating radiation. Essentially, no beta radiation was 

detected. No significant neutron exposures were registered by the network of activation foils in 

the access corridor. No evidence of internal contamination from inhalation was found. The 

absence of significant exposures is attributable to the extensive shielding provided by the process 

cell in which the event took place, and to the control of the fission gases by the equipment (Paulus, 

undated).

• Alpha-Contaminated Smear Samples, November 12, 1972: Routine health physics smear surveys 

of the Mass Spectrometry Laboratory (CPP-602, Labs 212 and 216) revealed low-level alpha 

contamination. The contamination on the high-level surfaces with predominant downward air 

flow indicated a strong possibility of significant air concentration. There were no air samplers in 

operation at the time. Therefore, whole-body counting, fecal, and urine analyses were performed 

for the individuals involved. On November 9 and 10, neptunium-plutonium samples prepared as 

nitrates contained about 0.5 g/1 of neptunium and about 0.07 g/1 of total plutonium, with the 

following mass percentages of plutonium isotopes: Pu-238, 5 w%; Pu-239, 80 w%; Pu-240, 10 

w%; Pu-241, 2 w%; and Pu-242, 1/2 w%. The activity of each sample was estimated at about 30 

μCi. These samples were normally in the range of 6-10 μCi. The neptunium was separated from 

the plutonium fraction by wet-chemical means within an alpha-tight glove box located in X-cell. 

Minor internal plutonium contamination of thirteen analytical laboratory personnel occurred 

during what should have been a routine sample-preparation procedure. A contributory cause was 

an abnormally-high quantity of plutonium in one sample. Resulting lung doses were small 

fractions (less than 2%) of the AEC guidelines, but the intensive bioassay-sampling program 

employed revealed an apparent previous plutonium exposure to [number of people redacted] of 

approximately one-third of the maximum-permissible lung burden for Pu-239 (Wenzel, 1973). It 

was determined that this earlier exposure occurred around May 1, 1972, during the clean-up of XCell (Wenzel, 1973b). NIOSH review of this earlier exposure indicates that the activities 

conducted in X-Cell were documented in Health Physics log books, along with survey results 

indicating significant alpha contamination within the cell (HP Log Sheets, 1972). However, the 

log books do not indicate that these activities prompted a special bioassay. Because the [number 

of people redacted] exposed in this case was not a frequent occupier of X-Cell, [number of people 

redacted] would not have been included in the routine sampling program for this area (Stroschein, 

1967) 
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Another Pu-238 filament was run on November 17, 1972, and at the end of the day, the carton 

containing the used filaments was prepared for disposal. While performing this operation, the 

bottom fell out of the container, dropping the filaments tested that day onto the floor. HP was 

called, the filaments picked up, and the floor wiped up. Smear surveys subsequent to the clean-up 

still read 500 cpm; the local area was isolated and cleaned up on November 20, the following 

Monday (Malmstrom, 1972).

Health Physics procedures in place at the time of the incident (dated November 1968) specified 

that Labs 212-216 were to be smear-surveyed weekly. Review of Health Physics records revealed 

that, for the first seven months of 1972, the laboratories were smeared an average of once per 

week. Since August 1972, the frequency of the routine smear surveys was about once every three 

weeks, with intervals of up to six weeks between surveys (Malmstrom, 1972).

As a result of these findings, NIOSH has determined that it cannot demonstrate that all 

potential plutonium exposures at CPP were identified through normal radiological controls. 

• Alpha Spill in S-Cell, February 1973: There was an alpha spill behind the S-Cell instrument panel. 

No detectable personnel exposures were recorded (Condotta, 1981). 

• Plutonium Spill in Y-Cell, November, 1973: Detectable personnel exposures were below 

maximum permissible dose (Condotta, 1981).

• Plutonium Spill in Access Corridor, January 1974: No significant personnel exposures recorded 

(Condotta, 1981). 

• Spill of Sample from Neptunium Recovery Process, September, 1974: The sample spill occurred 

in CPP-602, Rooms 204, 207, and 209. No detectable personnel exposures recorded (Condotta, 

1981).

• Plutonium Spill in CPP-602, February 1975: A plutonium spill occurred in CPP-602, Room 703. 

No detectable personnel exposures (Condotta, 1981).

5.4.3 Incidents at Test Area North

• Uranium Contamination Found in TAN-615, March 1963: In March 1963, TAN-615 was handed 

over to the Phillips Petroleum Company (Phillips), which had the STEP contract. Upon turnover, 

Phillips performed a thorough inspection and unexpectedly found the entire building to be highly 

contaminated with alpha radioactivity (TAN HP Report, Dec1963, PDF pp. 45-49). The alpha 

radioactivity was later determined to be from uranium (Dierks, 1963). 

• Alpha Contamination Found in Fuel Storage Vaults of TAN-607 Building, February, 1963: In 

February 1963, the Fuel Storage Vaults in TAN-607 were found to be “contaminated to a fairly 

high level with alpha contamination” and were decontaminated during that same month (TAN HP 

Report, Dec1963, PDF pp. 50-54). The spot with the highest contamination was in the bottom of a 

file cabinet drawer that was full of coffee grounds (TAN HP Report, Dec1963, PDF pp. 50-54). 
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Given that only uranium-based fuels were used in the TAN reactors, the alpha contamination in 

the Fuel Storage Vaults was likely limited to uranium contamination. 

5.4.4 Incidents at the Misc. Reactor Areas

• Stationary Low-Power Reactor #1 (SL-1) Reactor Explosion, January 3, 1961: A three-member 

military operating crew was involved in a nuclear incident at SL-1, a prototype nuclear power 

plant designed to provide power and heat for remote military installations. During reactor 

maintenance, one member of the military crew withdrew the central control rod well beyond the 

limit specified in the maintenance procedure. This resulted in a release of nuclear energy and the 

rapid formation of steam in the pressure vessel, accelerating a column of water above the core and 

slamming it into the pressure vessel. The impact of the compressed water sheared the connecting 

piping, lifted the vessel into the air, and contaminated the reactor building. Although the reactor 

building confined most of the radioactive material, about 1,100 curies of radioactive material were 

released into the atmosphere. This external release, according to the DOE historical dose 

evaluation study, did not expose the public to radiation above the current standard. All three 

members of the military crew were killed by the force of the reactor explosion or by injuries 

related to the explosion. Two were killed in the blast itself, and the third died two hours later from 

a head injury. In addition, 22 individuals subsequently involved in retrieving the bodies and 

cleaning up the site received radiation exposures ranging from 3 to 27 rem. A total of nine 

individuals received radiation doses exceeding the 15-rem standard in effect in 1961. However, 

14 individuals received radiation doses that exceeded the current 5-rem standard (SL-1 IDO 

Report, 1962).

An extensive personnel monitoring program was implemented immediately following the SL-1 

accident and continued throughout the recovery effort. A single entry point was established to 

ensure that the required personal protective equipment was used, that film badges were worn, and 

that briefings were provided to minimize work time in high-radiation and contamination areas. On 

January 5, 1961, IDO Personnel Metering Branch began issuing a daily report that listed worker 

names, dates, and amounts of radiation exposure attributable to the SL-1 incident (IDO Report, 

1961). Personnel Exposure Questionnaires (badge reports) were established for those initially 

involved in the incident response. These reports were used when an employee’s exposure reached 

a total of 1000 mR or more during the month (SL-1 Records, 2009). Examples of Health Physics 

Data reports, which include personnel exposures and information about the number of entries into 

SL-1, are available (Hoover, 1961; Layfield, 1961). 

Following the incident, personnel were exposed to airborne radioactive particles. Spot urine 

samples were collected from 110 individuals during the first few days after the incident (IDO 

Report, 1961). Thyroid counting was also implemented because I-131 was the major fission 

product of concern. All personnel involved in the SL-1 recovery effort submitted urine samples 

monthly or as needed based on radiological conditions. All samples were processed by the IDO 

Health and Safety Division. Urine samples were placed in a deep-well scintillation counter and 

gamma- counted. Gamma spectra indicated that essentially all of the radioactive material in the 

urine was I-131. Samples with statistically-significant activity underwent radiochemical 

separation and were beta-counted. Eight workers had urinary excretion of Sr-90 beyond Day 10 

post-exposure and started 24-hour urine samples to establish urinary excretion curves. Repeated 

samples were taken as needed and whole-body counts were done on all individuals with positive 
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gamma results (SL-1 Final, 1962). Six persons involved in the SL-1 incident reported to wholebody counting during the first week following the incident. Identification of I-131, Ba-140/La140, and Cs-137 was made by gamma spectra from all six of these workers (IDO Report, 1961). 

5.4.5 Incidents at Central Facilities Area

• For the purpose of this SEC evaluation report, there were no significant incidents to report at the 

Central Facilities Area.

5.4.6 Incidents at the Burial Ground

• “Chinook” Flood Event at the Burial Ground, February 1962: Heavy rainfall on top of existing 

snow caused a water run-off event from the surrounding area into the Burial Ground. Pits 2 and 3, 

as well as Trenches 24 and 25, were open at the time and filled with water. Breakage of some 

waste boxes allowed contaminated items such as gloves and bottles to be distributed in 

undeveloped areas adjacent to the Burial Ground. All detectable contamination was confined to 

the local areas. The event did initiate the installation of a diversion drain system around the 

perimeter of the Burial Ground (RWMC History, 1979).

• Fires in Burial Ground Trench 42, September 1966: On September 8 and 9, 1966, fires originated 

from waste buried in cardboard boxes. The cause of the fires was an inadvertent inclusion of 

alkali metals with low-level waste. The AEC-ID Fire Department extinguished the fires with 

water and by bulldozing soil over the items. Neither property damage nor detectable spread of 

contamination resulted from the event. During that time, it was decided to increase the minimum 

soil cover of buried waste from 0.6 to 0.9 meters and to increase the minimum trench depth from 

0.9 to 1.5 meters (RWMC History, 1985).

• Flood at Burial Ground, January 1969: During a two-day thaw, rainfall combined with melting 

snow flooded the Burial Ground. The flooding of Pits 9 and 10 as well as Trenches 48 and 49 was 

partially caused by large snow drifts that blocked the drainage system installed after the 1962 

“Chinook” flood. The water runoff overflowed the old dikes and flowed into the Burial Ground. 

As a result of the 1969 flood, new dikes and ditches were installed so that heavy equipment could 

be used to clear snow drifts. No significant spread of contamination resulted from this flood event 

(RWMC History, 1979).

• Fire at Burial Ground Temporary Storage, June 1970: A fire, started by sunlight shining on a black 

drum containing depleted uranium turnings and a loose lid, was discovered by security personnel 

at an above-ground temporary storage location. Initial attempts to extinguish the fire by the Fire 

Department were unsuccessful. An equipment operator used a crane to isolate the drum so that a 

bulldozer could cover the drum with soil to extinguish the fire. Air and direct radiation readings 

were constantly monitored with “very low” contamination reported (Fire, 1970).
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6.0 Summary of Available Monitoring Data for the Class Evaluated 

by NIOSH

Comprehensiveness of Individual INL Monitoring Data Received from DOE

When an individual who worked at INL files a claim under EEOICPA, the DOE responds with 

monitoring data for that individual. With one exception, NIOSH has not seen any indication that the 

DOE responses do not include all of an INL worker’s internal and external dosimetry records. In 

2012, personnel from the INL’s EEOICPA office made NIOSH aware that, in the past (period 

currently undefined), monitored visitors with no positive external doses and no bioassay data would 

not be entered into the INL’s dosimetry system, and the DOE responses received by NIOSH would 

not contain the external dosimetry records for those visitors. NIOSH was also informed that the INL 

site does have the external dosimetry records for those visitors; however, they are not in a readilyretrievable format for EEOICPA claims. Because the measured external doses for those visitors were 

all zero or less than their limit of detection, their missed external doses can be bounded by assuming 

the maximum dosimeter exchange frequency for a given year of monitored employment. This is not 

an issue when a worker has bioassay data because all workers with bioassay data were entered into the 

INL’s dosimetry program.

Comprehensiveness of INL Reports Received from DOE

NIOSH has not been able to verify that DOE responses provide all of the incident reports for workers 

involved with incidents. Incident reports provide useful information, such as intake date and source 

term data that aid in producing more accurate dose reconstructions. Incident information is 

periodically received in the DOE responses for INL claims; however, in 2014 and 2015, NIOSH 

captured a significant amount of incident records from the INL site and Seattle-NARA (National 

Archives and Records Administration). These records are being added to ORAUT’s Site Research 

Database (SRDB) and are being coded so that the names of the personnel identified in those reports 

will be linked to the EEOICPA claim as a “Personnel Exposure” file. NIOSH will continue to 

investigate, on a case-by-case basis, situations in which DOE responses for claims may be missing 

incident report information (based on the holdings in the SRDB). 

The following subsections provide an overview of the state of the available internal and external 

monitoring data for the INL class under evaluation.

6.1 Available INL Internal Monitoring Data

The Health and Safety Division of AEC’s Idaho Operations furnished a comprehensive personnel 

metering program for radiation monitoring of all NRTS personnel. The Analysis Branch of the 

division initially provided routine urinalysis of all NRTS personnel with sampling frequency varying 

between quarterly and yearly, depending on the expected potential for uptake (Horan, 1959). At the 

discretion of the chief health physicist of an individual program, the sampling frequency could be 

increased based on a perceived risk of uptake potential. For example, in the early 1960s, monthly 

24-hour urine samples were collected from all personnel who frequented X-Cell in CPP, including the 

HP technician. These samples were routinely processed for gross alpha, beta, gamma, and plutonium. 
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Any radioactivity detected by gross analysis was identified, and additional samples were requested 

when it was thought to be necessary (Stroschein, 1967).

Historically, INL did not routinely collect bioassay samples for uranium exposures although they 

could equate gross beta urinalysis results to uranium intake using an activity-to-mass conversion if 

there was a known uranium release; NIOSH has not found evidence of any such incident. INL 

controlled exposure to airborne uranium by use of continuous alpha and beta air monitoring. INL also 

used radiation survey smear data as an indication of uranium releases (Personal Communication, 

2014i); if an indication was found, bioassay samples would be collected from affected workers. 

INL did not routinely monitor workers for internal exposures to thorium. INL controlled exposure to 

airborne uranium by use of continuous alpha and beta air monitoring. INL also used radiation survey 

smear data as an indication of thorium releases (Personal Communication, 2014i); if an indication was 

found, bioassay samples would be collected from affected workers. Workers with potential for 

Pa-233 intakes were monitored using urinalysis and whole-body counting. 

Site-wide routine samples were initially analyzed only for gross alpha and beta, with isotopic analysis 

performed only if activity was discovered in the sample, although capabilities existed for analyses of 

specific isotopes, such as Pu-239, Sr-90, Co-60, and I-131 in body excreta, water, and air samples. 

The Analysis Branch also made analyses for materials related to good industrial hygiene practices, 

such as beryllium, lead, mercury, arsenic, fluorine, or any other chemical that may have been required

(Horan, 1959). Beginning in May 1957, the Health and Safety Division started requesting grossgamma counts on all urine samples submitted by CPP personnel in an effort to detect the potential 

uptake of iodine from fresh fuel (HP Monthly Reports, 1957). Table 6-1 provides INL in vitro

measurements for 1953-1975. 

Based on the loss of iodine due to volatilization during the wet-ashing process, as well as the small 

likelihood of a significant uptake of a pure beta-emitter without a correspondingly-large uptake of 

gamma-emitting fission products, the gross-beta procedure on urine samples was discontinued entirely 

in December 1960. However, in order to detect a slow build-up of the bone-seeking, beta-emitting 

strontium isotopes from low-level sources over long periods of time, specific strontium analyses were 

made for each individual that had worked in a reactor or chemical processing area for two years or 

longer (Murphy, 1966) every two years and on termination of employment (Horan, 1961).
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Table 6-1: In-Vitro Measurements for INL, 1953-1975

Year Gross 

Alpha

Gross Beta or 

Beta/Gamma

Gross 

Gamma

Cs Sr-89 / 

Sr-90

U-233 / 234 / 

235 / 238

Pu-238 / 

239 / 240

Total 

Urine (a)

Total 

Fecal

1953 - 455 - - - 1 - 456 -

1954 6 1400 2 - - 52 - 1466 -

1955 - 2774 1 - - 42 3 2896 -

1956 6 4924 1 - 89 17 - 5054 -

1957 - 7572 242 1 1 95 2 9798 -

1958 1 7398 240 - 4 133 3 10719 -

1959 1 7010 1048 - 37 57 21 9682 -

1960 - 6213 1794 - 69 40 9 8932 -

1961 - 40 7006 39 2096 2 8 9219 27

1962 8 8 6125 2 1579 13 18 7775 -

1963 - 19 5890 - 937 124 135 7275 75

1964 1 2 5401 - 2228 60 40 7763 33

1965 - 10 5610 3 1079 50 58 6871 6

1966 1 - 1823 - 1099 11 11 2956 3

1967 - 20 414 - 395 17 31 851 42

1968 - 447 302 - 328 - 15 1138 -

1969 - 148 581 2 182 1 19 940 -

1970 1 5 189 1 10 18 6 274 -

1971 - - 3 - - - - 5 -

1972 - - - - - - 29 58 335

1973 - - - - - - 58 58 103

1974 - - 1 12 13 - 35 76 81

1975 - 22 7 5 38 - 8 80 171

Source: Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986

a

 Total number of urine samples includes infrequently-sampled radionuclides not included in this table, such as H-3, 

Co-60, and others. 

Whole-body counting (WBC) was started at NRTS in January 1961. The convenience, sensitivity, 

and reliability of whole-body counting made it the choice of personnel monitors for internal 

radioactivity from gamma-emitting nuclides. Consequently, urinalysis as a routine monitoring 

technique was abandoned by the laboratory. From that time, urine was analyzed only for iodine, 

uranium, and strontium or for elimination studies on specific isotopes (Voelz, 1970). Early in the 

site’s history, job duties and their corresponding exposure potentials were used to place employees 

into the following routine WBC frequencies: quarterly, yearly, or no count needed. Special analyses 

were requested by the relevant Health Physics sections when internal exposure was suspected due to 

events occurring in a plant; terminations that required a physical examination, as determined by the 

relevant Health & Safety sections, also required a whole-body count (McCaslin, 1963). 

A portable, unshielded whole-body counter was developed and constructed in 1963. NOTE: This 

counter was used to monitor people involved in an incident that occurred when a shipment of methyl 

iodide released activity in an airplane. A medical van containing a whole-body counter was designed 

and purchased in 1965. Near the end of 1966, policy shifted to performing routine WBCs on 

employees once every four years, while maintaining the requirement to perform special analyses on 
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those involved in radiological incidents (McCaslin, 1966). By 1969, whole-body counting equipment 

available at the Analytical Chemistry Branch Laboratory included rotational counting, helical 

scanning, and a detector for determining Sr-90 in the skeleton by counting the Bremsstrahlung 

emission from the skull (Voelz, 1970). Table 6-2 provides INL in vivo measurements for 1961-1975. 

Table 6-2: In-Vivo Measurements for INL, 1961-1975

Year Number of Results

1961 1344

1962 2667

1963 3839

1964 1998

1965 2646

1966 3072

1967 1511

1968 1534

1969 1158

1970 1350

1971 1210

1972 1396

1973 1285

1974 794

1975 1626

Source: Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986

In May 1967, the Health Services Laboratory was organized from six branches of the Idaho 

Operations' Health and Safety Division. Its primary function was to develop a technical staff in 

occupational medicine, health physics, radiological dosimetry, analytical chemistry, and 

instrumentation to support the NRTS occupational health programs (Voelz, 1969). 

Although the analyses were performed at the HSL Analytical Chemistry Branch, over time, the 

requirements for determining which employees would be sampled and on what frequency became 

fragmented by department and facility. By 1974, bioassay samples were collected and analyzed only 

when an exposure incident was suspected. The typical problems were encountered in obtaining the 

cooperation necessary from exposed personnel (e.g., providing samples as requested and furnishing all 

necessary information). A routine program existed at that time for WBC measurements of CPP 

personnel for gamma-emitter uptake. Each individual working at CPP received a whole-body count 

every four years (i.e., 25 percent of the work force was counted every year). The majority of the 

counting was done using the HSL mobile WBC (NaI detector) that was brought to the CPP. The 

remaining counting was accomplished using the WBC facilities at the HSL (Rich, 1974).

Also during 1974, a routine bioassay program was being developed. The proposed program 

recommended the collection and analysis of about 350 bioassay samples per year. Of these, 50 were 

to be fecal samples to be analyzed for plutonium and/or uranium, 50 were to be urine samples to be 

analyzed for plutonium, and the remaining 250 were to be urine samples to be analyzed for total 
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uranium and fission products. The challenge at the time was that the HSL staff level was not 

sufficient to provide for this increase in routine bioassay sampling (Rich, 1974). 

Surveys, air samples, and bioassay samples for fission and activation products (via gross alpha and 

gross beta/gamma analysis) were routinely performed over the entire period under evaluation, as 

indicated by many available documents and a database (Bioassay, 1956-1984; Bioassay, 1956-1967). 

Workers were routinely assayed, by either urinalysis or whole-body counts, for potential intakes of 

fission and activation products and for alpha-emitting radionuclides (including but not limited to 

plutonium) when alpha contamination or air monitoring data exceeded background, or after the 

occurrence of an incident involving the airborne release of radioactive material, or when alpha 

contamination or air monitoring data exceeded background. 

Examples of radiological survey data can be found in the following examples: Survey, Jun1958; 

Survey, Nov1958; Survey, Oct1955; Survey, May-Jun1955; Survey, Aug-Dec1953; Survey, Apr1963; 

Survey, Aug1968; Survey, Jul1963; Survey, Apr1969. 

Details regarding the various analyses used and the associated minimum detectable activities are 

presented in the Technical Basis Document, Idaho National Laboratory and Argonne National 

Laboratory-West - Occupational Internal Dose (ORAUT-TKBS-0007-5).

6.2 Available INL External Monitoring Data

Personal dosimeter data (i.e., film badge and/or TLD) are the primary data type used to reconstruct the 

external doses for INL workers. Those data are specifically used to reconstruct a worker’s measured 

and missed external doses. For the period under evaluation for this report, the INL site monitored all 

workers for external photon and beta (electron) doses via their own dosimeters when inside the major 

radiological operating areas (Cipperley, 1958; CPP HP Manual, 1952; Cipperley, 1968; ORAUTTKBS-0007-6). The only known exception to this was for some construction workers. For some 

construction activities, parts of an operational area were fenced off and the construction workers 

inside those fenced areas were monitored by group dosimeters or area dosimeters. NIOSH does not 

possess a complete set of the INL’s dosimeter data; however, the DOE does have a complete set and 

provides the external dosimetry records for each energy employee who files an EEOICPA claim. A 

summary of the dosimetry data is provided in Table 6-3. 

Examples of survey data have been located but are not complete for all locations for all time periods 

(Survey, Jun1958; Survey, Nov1958; Survey, Oct1955; Survey, May-Jun1955; Survey, AugDec1953; Survey, Apr1963; Survey, Aug1968; Survey, Jul1963; Survey, Apr1969). 
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Table 6-3: Historic External Radiation Exposure Distribution for INLa

Year Dose Groupings (rem)b Total

Monitoredc

Site

Populationb 0-1 1-2 2-3 3-4 4-5 >5

1949 NA NA NA NA NA NA NA NA

1950 NA NA NA NA NA NA NA NA

1951 NA NA NA NA NA NA NA 2,250

1952 83 0 0 0 0 0 83 2,350

1953 738 34 3 2 2 1 780 2,175

1954 797 79 26 8 1 2 913 2,525

1955 1,336 120 76 29 10 4 1,575 3,025

1956 964 125 61 24 9 11 1,194 3,625

1957 1,243 114 20 4 0 0 1,381 4,550

1958 2,785 260 143 47 24 29 3,288 5,000

1959 2,962 257 196 33 14 5 3,467 5,075

1960 4,020 220 94 51 11 1 4,397 5,200

1961 3,199 346 201 138 45 19 3,948 4,920

1962 3,137 229 102 73 49 23 3,613 4,755

1963 3,283 212 87 77 23 0 3,682 5,310

1964 2,433 190 115 82 15 0 2,835 5,900

1965 2,547 439 221 180 144 61 3,592 6,025

1966 2,400 206 105 83 44 0 2,838 5,800

1967 2,502 226 109 95 53 0 2,985 5,825

1968 2,612 163 100 86 44 1 3,006 5,850

1969 2,526 196 81 70 23 0 2,896 5,850

1970 2,936 0 168 77 59 33 3,273 5,755

1971 2,723 136 57 27 4 0 2,947 5,680

1972 2,553 171 89 59 19 0 2,891 5,875

1973 2,143 183 66 36 18 1 2,447 5,975

1974 2,774 133 62 20 1 0 2,990 5,950

1975 2,399 115 51 15 0 0 2,580 3,643

Source: Exposure History, 1993, PDF p. 41

Notes:

NA = Not available

a - The numbers presented in this table are inclusive of AEC and contractor personnel engaged only on contracts 

administered by IDO; they do not include visitors or personnel employed on contracts administered by other AEC 

operations offices. 

b - Not all of these numbers are consistent with the numbers reported in other references, such as the available Health & 

Safety Annual Reports (Horan, 1961, PDF p. 31; Horan, 1962, PDF p. 62). After reviewing the available INL 

documentation, NIOSH has concluded that definitive historical site population numbers are elusive. There is no 

consistent source of employment information and the various sources count employees differently. INL’s own

historical overview of site exposure history points out that there is little evidence of real effort to identify and preserve

historic data. Those INL authors experienced considerable difficulty obtaining total counts before 1974, much less a 

breakdown by individual contractor (Exposure History, 1993). Most sources count only the AEC and contractor 

personnel administered by IDO; others may count non-IDO contract employees (e.g., ANL-W), and sometimes 

visitors are counted. These counts do not appear to include workers employed by minor subcontractors (e.g., 

construction workers), which only appear to be included when visitors are counted. Counting visitors can greatly 

expand the population count (e.g., for 1961, from 4,920 to 25,580) (Horan, 1962, PDF p. 61). Thus, INL 

employment figures are best viewed as “ballpark” numbers that reflect AEC and primary contractor personnel.

c - These numbers are the sums from the dose groupings for a given year. In addition, they do not include visitors or 

personnel employed on contracts administered by other AEC operations offices, which could be significant (see Note 

b above).
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Because all the workers in INL’s radiological areas were monitored for external dose, and because the 

DOE provides each worker’s dosimeter data for NIOSH dose reconstruction, the only doses that need 

to be assigned for the unmonitored workers at INL are onsite ambient (environmental) external doses. 

Because the control dosimeters at INL were likely exposed to elevated levels of onsite ambient 

radiation due to being located at the entrances of the major operating areas, onsite ambient doses are 

also assigned to all monitored INL workers to account for any radiation dose that may have been 

inappropriately subtracted from their dosimeter results as background radiation (ORAUT-TKBS0007-6).

For the period under evaluation for this report, NIOSH has environmental dosimeter data for 1973 and 

beyond. The environmental dosimeters associated with those data were located along the security 

fence lines for each major operating area. Because the closest that an unmonitored worker could get 

to a major operating area was the security fence lines for those areas, the onsite ambient doses at those 

fence lines are considered to be bounding doses for unmonitored workers. A summary of the annual 

environmental doses for each major operating area is provided in ORAUT-TKBS-0007-4, Table 4-13. 

For the years prior to 1973, the highest 6-month value from April 1972 to April 1973 for an operating 

area was multiplied by two and applied to each year between 1952 and 1972. A more detailed 

discussion of the onsite ambient doses used for INL dose reconstructions can be found in of ORAUTTKBS-0007-4, Section 4.3.1. For the years 1949–1951, the onsite ambient doses can be bounded by 

assigning the onsite ambient doses being assigned for the years 1952–1972 because radiological 

operations prior to 1952 were minimal compared to 1952 and later (ORAUT-TKBS-0007-2). To 

bound any onsite ambient dose when an energy employee’s work location for a given year was 

unknown (often the case for unmonitored workers), the highest onsite ambient dose for any operating 

area can be assigned for that year. 

Occupational medical doses for INL workers can be assessed based on X-ray records when provided 

for the EEOICPA claim, which is now common for new INL claims. When occupational X-ray 

records are not available for an INL worker, the occupational medical doses can be estimated using 

the site-specific information about the examination frequencies and types of exposures in ORAUTTKBS-0007-3. ORAUT-TKBS-0007-3 also provides the organ doses for the X-ray procedures that 

were performed at INL.
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6.3 Available Radiological Monitoring Data

Consideration of personnel exposures to radioactive material commenced before the start of 

radiological operations at Idaho National Laboratory. INL took advantage of the practices and 

experiences of other AEC facilities established earlier, such as Oak Ridge National Laboratory and 

Hanford, to establish their health physics program. AEC-IDO policy regarding health physics 

responsibilities and services at INL was established in 1952 creating centralized services (Johnston, 

1952). INL used air sampling and radiological contamination monitoring programs as qualitative 

indicators of internal exposures, with routine bioassay programs serving as verification programs that 

internal exposures had not occurred. 

The contamination control limits for the detection and control of released activity beyond the control 

boundaries related to instrumentation capabilities and the basic philosophy of acceptance of detectable 

contamination. The contamination control limits for on-plant surfaces, and particularly personnel, 

were always close to the MDA, so that any detectable contamination was a signal for follow-up 

evaluations and actions.

The monitoring of radioactivity in the air in occupied areas was a fundamental component of the 

internal exposure control program. Beta/gamma CAMs were used from the beginning of all facility 

and program operations in routinely-occupied areas. The CAM systems provided real-time 

air-activity evaluations while fixed air samplers provided retrospective data and an average air 

concentration in an area or building. 

6.3.1 Available Radiological Monitoring Data for TRA 

From the onset of radiological operations through the end of evaluation period, TRA Health Physics 

conducted routine continuous air sampling at MTR, ETR, and other areas with the potential for release 

of airborne radioactivity. Air monitoring checked for fission products and alpha releases. CAM 

filters were counted for long-lived beta-gamma and alpha activities later after collection (CAM, 

1961). NIOSH has collected a sampling of TRA air sampling data; data for all the years under 

evaluation are available at INL. A comprehensive radiological survey program existed at TRA. 

Figure 6-1 shows an example radiation survey (ETR Surveys, 1961-1962); Figure 6-2 shows an 

example smear survey (Plutonium in Feces, 1965).
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Figure 6-1: Radiation Survey of KAPL F/H-10 Cubicle in ETR, 4/24/64
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Figure 6-2: Radiological Survey of MTR Alpha Labs June 6, 1962

6.3.2 Available Radiological Monitoring Data for CPP 

A review of survey and monitoring data from 1955-1971 suggests a changing attitude toward 

contamination control. While contamination in the CPP-601 Access Corridor was noted on survey 

logs in the early years of operation, routine operations required that contaminated areas be mopped to 

reduce the contamination level to below guidance levels. This same precedent is seen in the 

fluctuating contamination levels in the Operating and Sample Corridors, where one month shows 

contamination spread between B- and H-Cells and a survey from the following month shows no such 

contamination (CPP Routine HP Surveys with Alpha Results 6/29/61 – not yet in SRDB)

Contamination potential existed from each of the cells, which were highly contaminated. The greatest 

contamination potential existed as a result of spills, or during plant shutdown periods when the 

centrifuge shielding was removed or cells were entered. However, in later years, general levels in the 

corridors during normal operations routinely required protective clothing and contamination control 

barriers. The Access Corridor was contaminated routinely to several thousand dpm/100 cm2 by the 

1970s and the area was designated a Shoe Cover Area necessitating the wearing of protective 

equipment. Figures 6-3 through 6-7 successively show a comparison of contamination surveys in the 

CPP-601 Access Corridor area between 1961 and 1970. Note the presence of a Shoe Cover (SC) Area 

in the Access Corridor in Figure 6-7 (CPP HP Surveys, 1970; CPP Routine HP Surveys with Alpha 

Results 6/29/61 – not yet in SRDB).
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Note: “RCG” in the above image refers to Radiological Control Guidelines.

Figure 6-3: Comparison of CPP-601 Access Corridor Contamination Surveys: 1961
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Figure 6-4: Comparison of CPP-601 Access Corridor Contamination Surveys: 1963
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Figure 6-5: Comparison of CPP-601 Access Corridor Contamination Surveys: 1965
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Figure 6-6: Comparison of CPP-601 Access Corridor Contamination Surveys: 1969 
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Note: “SC” in the above image refers to a Shoe Cover Area.

Figure 6-7: Comparison of CPP-601 Access Corridor Contamination Surveys: 1970
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Significant levels of Pu contamination (>10 dpm/ 100 cm2) were identified in the early 1970s in a 

number of cells (Rich, 1974).

Procedurally, all routine area surveys should have been spot-checked for alpha by alpha-counting 

approximately one-fifth of all smears taken. In surveys of areas where alpha-emitters were normally 

found or expected, all smears were to be counted for alpha contamination. Documentation of 

incidents and review of the survey data available to NIOSH suggests that personnel did not strictly 

adhere to this survey frequency (Malmstrom, 1972). 

A review of CPP’s incident history indicates a long record of repeated contamination incidents and 

temporary fixes. It is apparent that less-than-adequate contamination-control conditions existed at the 

CPP for many years - the shift lab being one of the worst examples. Several investigating committees 

made comprehensive recommendations to correct the conditions and prevent reoccurrences (Rich, 

1974). Presumably, due to the difficulty of re-engineering an existing facility, the difficulty of 

changing worker habit patterns, and the anticipated loss in productivity, these recommendations were 

not implemented (Condotta, 1981a). 

Events involving uncontrolled contamination on personnel and vehicles at CPP-603 in 1970 

emphasized the problems there. All surfaces in the building were contaminated to levels of 103 to 105

dpm/100 cm2

, and some areas and equipment were so highly contaminated that it was impossible to 

avoid routine personnel contamination (Rich, 1974a). Health Physics procedures specified that Labs 

212-216 were to be surveyed by smearing on a weekly basis. Review of Health Physics records and 

survey log sheets revealed that, for the first seven months of 1972, the laboratories were smeared an 

average of once per week. Between August and December of that year, the frequency of the routine 

smear surveys was about once every three weeks, with intervals of up to six weeks between surveys 

(Malmstrom, 1972).

Figure 6-8 presents an example contamination survey for CPP-204 and CPP-207 in 1968. Note the 

alpha contamination on open bench tops and a desk (Surveys, 1968).
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Figure 6-8: Example Contamination Survey for CPP-204 and CPP-207 in 1968
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6.3.3 Available Radiological Monitoring Data for TAN 

Air Monitoring

The various Health Physics organizations at TAN appear to have monitored the airborne radioactivity 

for several locations throughout TAN. NIOSH has collected a sampling of TAN air monitoring data 

for the years 1962–1970 (CAM 1967-1970; CAM, 1962-1963; CAM, 1963-1964; CAM, 1964-1967). 

A number of ANP Program records indicate that General Electric (GE) monitored airborne 

radioactivity during the ANP Program era (i.e., prior to 1962) (D101 L2A-1, 1959; Boone, 1959; 

Levine, 1959); however, NIOSH has not located any of the results for those samples. A 1982 

memorandum from GE indicates that the radiological records from the ANP Program are not at the 

INL site and were likely transferred to GE’s Evendale facility when GE’s contract with the IDO-AEC 

terminated (Robertson, 1982). 

The available information indicates that airborne radioactivity was monitored using a combination of 

CAMs, portable air samplers, and fixed head air samplers (referred to as “vacuum air system 

locations” in the records) (D101 L2A-1, 1959; CAM 1967-1970; CAM, 1962-1963; CAM, 1963-

1964; CAM, 1964-1967). During the ANP Program, air samples may have only been analyzed for 

gross beta-gamma radioactivity (D101 L2A-1, 1959). After the ANP Program, air samples were 

analyzed for gross beta-gamma radioactivity and gross alpha radioactivity (D101 L2A-1, 1959; CAM 

1967-1970; CAM, 1962-1963; CAM, 1963-1964; CAM, 1964-1967). 

Radiological Surveys

The various Health Physics organizations at TAN appear to have periodically performed radiological 

surveys throughout TAN (D101 L2A-1, 1959; TAN Monitoring, 1963). NIOSH has collected a 

sampling of the TAN radiological data for the years 1964–1968 (TAN Monitoring, 1963). ANP 

Program records indicate that GE performed radiological surveys during the ANP Program era (i.e., 

prior to 1962) (D101 L2A-1, 1959); however, NIOSH has not located any of the results for those 

surveys. A 1982 memorandum from GE indicates that the radiological records from the ANP 

Program are not at the INL site and were likely transferred to GE’s Evendale facility when GE’s 

contract with the IDO-AEC terminated (Robertson, 1982).

The available information indicates that contamination surveys were predominately for beta-gamma 

radioactivity, but alpha contamination surveys were periodically performed. Figure 6-9 is an example 

of a 1963 survey schedule for part of the TAN-607 Building (TAN Monitoring, 1963, PDF p. 105). 

Figure 6-10 is an example of a 1968 contamination survey of the TAN-633 Building (Hot Cell Annex) 

(TAN Monitoring, 1963, PDF p. 119). 
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Figure 6-9: Survey Schedule for Part of the TAN-607 Building
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Figure 6-10: Example of Contamination Survey of TAN-633 (Hot Cell Annex)
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6.3.4 Available Radiological Monitoring Data for Misc. Reactor Areas

Although the Auxiliary Reactor Area, Organic Moderated Reactor Experiment, and Special Power 

Excursion Reactor Test facilities were separate and distinct programs, all had sound health physics 

programs throughout their years of operations. Surveys and air monitoring were consistently and 

routinely performed, as indicated by monthly reports, activity release reports, and log books (Layfield, 

1961a; OMRE, 1959; HP Logbook, 1959). In addition, internal and external appraisals of health 

phyics programs were performed (OMRE HP Survey, 1962-1963). Figure 6-11 shows an excerpt 

from a SPERT Health Physics Monitoring Report for August 1963 (Terch,1963). 

Figure 6-11: Excerpt from SPERT Health Physics Monitoring Report, Aug. 1963
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6.3.5 Available Radiological Monitoring Data for CFA 

The Central Facilities Area was unique to INL in that CFA employees worked “outside of plant 

fences.” Radiation or contamination issues were typically encountered inside fenced areas. However, 

there were specific areas at CFA where safe work permits and protective clothing were required as 

precautions. Contamination surveys and air monitoring were routinely performed in these areas to 

monitor internal exposures (CFA HP Logs, 1965-1966). This monitoring was summarized in CFA 

Health Physics monthly reports. Figure 6-12 shows an excerpt from a CFA Health Physics Monthly 

Report for July 1961 (Johnson, 1961).

Figure 6-12: Excerpt from CFA Health Physics Monthly Report, Jul. 1961
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6.3.6 Available Radiological Monitoring Data for the Burial Ground

Safe Work Permits and monitoring practices at the Burial Ground (e.g., required air monitoring) 

indicate contemporary controls and enforcement of restricted areas and transportation vehicles. A 

health physicist was required to be present for waste disposal and also had overall responsibility for 

Burial Ground operations. Entry and exit monitoring served as the crux of contamination control for 

personnel and vehicle contamination control throughout the evaluation period. Figure 6-13 shows a 

Safe Work Permit for RFP waste burial at the INL Burial Ground for April 1964 (SWP 1962-1964) 

Figure 6-13: Safe Work Permit for RFP Waste Burial at INL Burial Ground, Apr. 1964
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7.0 Feasibility of Dose Reconstruction for the Class Evaluated by 

NIOSH

The feasibility determinations for the class of employees under evaluation in this report are governed 

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

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

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

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

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

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

conduct dose reconstructions. 

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

reconstructions demonstrate the feasibility of estimating with sufficient accuracy the potential 

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

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

which together or individually might assure that NIOSH can estimate either the maximum doses that 

members of the class might have incurred, or more precise quantities that reflect the variability of 

exposures experienced by groups or individual members of the class as summarized in Section 7.6. 

This approach is discussed in DCAS’s SEC Petition Evaluation Internal Procedures which are 

available at http://www.cdc.gov/niosh/ocas. The next four major subsections of this Evaluation 

Report examine: 

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

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

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

• The bases for petition SEC-00219 as submitted by the petitioner. (Section 7.4) 

7.1 Pedigree of INL Data

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

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

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

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

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

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

feasibility evaluation presupposes that data pedigree issues have been settled.
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7.1.1 Internal Monitoring Data Pedigree Review

Upon request, INL provides database printouts and copies of original hardcopy bioassay results, and 

whole-body-count/chest-count records. NIOSH has obtained both bioassay results and a limited 

number of air monitoring results for the period under evaluation. Bioassay sample results were 

obtained starting in 1953. The bioassay results are urinalyses starting in 1953; fecal sample results 

start in 1961. NIOSH has also obtained whole-body-counting results starting in 1961. 

A large amount of raw bioassay data exists at INL. Many of these records have been compiled into a 

single database with a data validation and comment field added to each record. These data are in the 

process of being cross-verified with scanned copies of 32,951 pages of internal dosimetry records 

(Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986). Duplicate records and unusable 

records (e.g., gross errors) are not included in this assessment.

There have been a number of "spot checks" performed to confirm the integrity of the in-vitro

radiobioassay dataset for INL. These spot checks took the form of: 

1. Capturing hard copy records of bioassay records and checking the results against the dataset

2. Having INL generate a listing of all monitoring records for 12 workers and checking the values 

against the dataset

These "spot checks" consisted of:

• Termination in-vitro results for 24 Fluor Corporation workers between August 15, 1958 and 

November 23, 1958 

• Routine and special in-vitro results for 95 workers in 1963 for gamma, Pu, Sr, and U analyses 

• All in-vitro records for 12 INL workers, including chemists believed to have been involved in Pu 

separations 

• CF-669 laundry workers, as identified from external dosimetry reports 

In all the cases above, the results were found in the INL in-vitro radiobioassay dataset or the PDFs 

used to create the dataset. An exhaustive internal data pedigree review of the databases has not been 

performed because of the existence of original hard-copy personnel records. Additional checks of the 

data in the databases will be performed and documented as part of the development of the internal 

dose co-worker studies that are planned for INL.
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7.1.2 External Monitoring Data Pedigree Review

Upon request, INL provides database printouts and copies of original hard-copy external dosimetry 

results. NIOSH has obtained personnel dosimetry information and area monitoring data for the period 

under evaluation. External dosimetry results were obtained starting in 1952. A large amount of 

personnel dosimetry data exists at INL, including duplicate records and unusable records (e.g., gross 

errors). NIOSH has determined that all the workers in INL’s radiological areas were monitored for 

external dose and the dose values exist in original hard-copy form as well as database form, which is 

(and has been) provided by DOE for each worker. In addition to the dosimetry data that are 

considered a primary resource for supporting external dose reconstructions, NIOSH also has access to 

INL area monitoring data (including radiation survey data and source-term information) as well as 

onsite ambient (environmental) data, which are also in original hard-copy form. These data support 

the assignment of onsite ambient doses for all monitored INL workers to account for any radiation 

dose that may have been inappropriately subtracted from their dosimeter results as background 

radiation, as well as the assignment of dose to unmonitored individuals.

7.2 Evaluation of Bounding Internal Radiation Doses at INL

The principal sources of internal radiation doses to members of the proposed class would have been 

inhalation and ingestion of radiological contamination during the following operations: 

• Destructive and non-destructive nuclear reactor tests

• Irradiation of experimental nuclear reactor fuels

• Irradiation of various materials to transform them by neutron bombardment and/or to study their 

behavior within a nuclear reactor environment 

• Chemical separation of highly-enriched uranium from irradiated fuel elements

• Chemical separation of neptunium from irradiated fuel elements 

• Research and development activities, including human and environmental studies 

• Waste management operations supporting all site radiological activities

The potential internal sources would have been dependent on the operational area and activities. The 

major sources of intakes, as indicated in the bioassay records and incident reports, have been mixed 

fission products (often limited to radioactive iodines and noble gases), activation products, uranium, 

and plutonium. However, the records list a wide spectrum of radionuclides that were monitored and 

an even longer list of codes used to identify the radionuclides, groups of radionuclides, specific 

measurement techniques, or combinations of radionuclides and techniques. Many of the radionuclides 

apply to a small set of workers on a research project or to workers whose tasks could have exposed 

them to many different sources (e.g., radiation monitoring technicians) (ORAUT-TKBS-0006-5). 
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To provide consistency to INL radiation safety programs given the large variety of facilities and 

constantly-changing contractors, the AEC established a Health and Safety (H&S) Laboratory at INL 

to provide technical support for internal and external dosimetry programs. The H&S Laboratory was 

operated by the AEC rather then one of its INL contractors. The name of this organization changed to 

Health Services Laboratory (HSL), then to the Health and Safety Division (HSD), then to the Idaho 

Center for Radiological and Environmental Sciences, and most recently to the Radiological and 

Environmental Sciences Laboratory (RESL). During the period under evaluation, in-vitro bioassay 

samples collected throughout INL were analyzed by the chemistry laboratory HSL, which was located 

in Building CF-646 in the CFA. In 1963, in-vitro bioassay analysis was moved to a new laboratory in 

Building CF-690. With the move to CFA-690 in 1963, the whole-body counting facilities were 

housed in the same buildings as the bioassay labs. A newly-designed counter was used from the onset 

of CFA-690 operations (Puphal, 1994).

For INL EEOICPA claims, ambient environmental internal doses are only assessed for certain 

unmonitored workers, as discussed in the guidance in ORAUT-TKBS-0007-6, Section 5.6. ORAUTTKBS-0007-4 and its supplement (Peterson, 2004) provide the historical background, rationale, and 

environmental intake data for reconstructing occupational environmental internal doses at the INL 

site. In summary, ORAUT-TKBS-0007-4 provides annual environmental intakes of the significant 

radionuclides for each major operating area on the INL site. Because environmental exposures are 

accounted for in personnel occupational exposures and do not impact the ability to reconstruct dose 

with sufficient accuracy, further assessment of environmental internal exposure is not included in this 

report.

The following subsections address the ability to bound internal doses, methods for bounding doses, 

and the feasibility of internal dose reconstruction for the six operating areas under evaluation. 
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7.2.1 Evaluation of Bounding Internal Doses for TRA 

7.2.1.1 Evaluation of Bounding Process-Related Internal Doses for TRA

The principal sources of internal radiation doses for members of the class working at the Test Reactor 

Area include plutonium and other transuranic radionuclides, uranium, thorium, and mixed fission 

products. The following subsections address the ability to bound internal doses, methods for 

bounding doses, and the feasibility of internal dose reconstruction. 

A tabular representation of the feasibility of reconstructing doses is presented in Table 7-1, where the 

dark green cells with an “F” indicate reconstructions that are feasible and the lighter green cells with a 

“C” indicate reconstructions that are feasible but a mixed-fission-product co-worker model is needed. 

The rationale for these feasibility designations is presented in the following subsections.

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible but a mixed-fission-product co-worker model is needed.

Urinalysis Information and Available Data 

As previously discussed in Section 7.1, there are extensive urine bioassay data that NIOSH believes 

are sufficient, along with whole-body count (in vivo) analysis results, for bounding internal radiation 

doses from plutonium, uranium, and mixed fission products. Sampling was done annually, quarterly, 

or as the Heath Physicist ordered, with more frequent sampling done following an incident. In the

1960s, fecal analyses were incorporated into the bioassay program to assess any possible lung burden 

from the alpha-emitting radionuclides. 

In the available data, the earliest urine bioassay data collected from TRA workers are for plutonium 

starting in 1955, uranium in 1955, transuranic radionuclides in 1959, and fission products in 1953 

(Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986). The earliest in vivo data for TRA 

workers in the available data are for 1961. The in vivo data will be used with urinalysis results to 

bound doses received from X-ray and gamma-emitting radiation. The number of samples by type and 

year are provided in Table 7-1. 

Table 7-1: Feasibility Summary for TRA (1952-1970)

Exposure Source

Years

52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70

Mixed Fission Products F F F F F F F F F F F F F F F C C C C 

Plutonium F F F F F F F F F F F F F F F F F F F 

Other F F F F F F F F F F F F F F F F F F F 

Photon F F F F F F F F F F F F F F F F F F F 

Neutron F F F F F F F F F F F F F F F F F F F 

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TRA controlled exposure to airborne and re-suspended radioactivity by continuous air monitoring 

(CAM) and smear survey programs. Examples of CAM are available (Air Monitoring, 1965-1966, 

Air Monitoring, Feb1953-Oct1954, Air Monitoring, Dec1956-Aug1957, Keller, 1958, Air 

Monitoring, 1966-1967, Air Monitoring, 1966, Air Monitoring, Mar1965-Dec1966). Examples of 

contamination survey data are available (Survey, Jun1958, Survey, Nov1958, Survey, Oct1955, 

Survey, May-Jun1955, Survey, Aug-Dec1953, Survey, Apr1963, Survey, Aug1968, Survey, Jul1963, 

Survey, Apr1969). During the period under evaluation, routine urinalysis and whole-body-count 

programs were conducted at TRA for intakes of fission and activation products (WBC1, 1961-1977; 

WBC2, 1963-1995; WBC3, 1981-1994; WBC4, 1979-1995; WBC5, 1973-1996). From 1961 until 

1966, TRA workers were given whole-body counts on a routine frequency based on their potential for 

exposure (Sommers, 1961; McCaslin, 1963). By 1965, whole-body counting became the analysis of 

choice to determine intakes of fission products. INL employed an incident-based bioassay program 

for detection and quantification of intakes of radioactive material. INL determined that whole-body 

counting provided the necessary detection sensitivity for most fission products, and in some cases, 

provided detection superior to urinalysis, even if using 24-hour urine sample (Puphal, 1994, PDF p. 

11). In 1967, INL changed whole-body-count frequency to randomly counting one-fourth of the 

workforce each year whereby each worker would be counted every four years (Sommers, 1966). A 

roster of workers to be counted in 1968 is available (Sommers, 1968).

Special samples were collected from workers when either results of CAM or smear surveys suggested 

internal exposure (McCaslin, 1963). As stated in Section 5.4.1, there were often releases of fission 

product radionuclides. Special urinalysis or in-vivo analyses were requested when incidents resulted 

in airborne radioactivity being detected. NIOSH has obtained bioassay for TRA incidents involving 

fission products and alpha-emitters. Results provided in Table 7-2 for plutonium, uranium, and 

transuranic radionuclides represent special samples collected following alpha-based incidents or 

where smear-survey results exceeded TRA guideline values. For the Pu-240 incident that occurred in 

TRA in 1963, 99 bioassay samples were collected from 78 workers exposed to the contamination. 

One sample from each of six people was statistically positive. Those results ranged from 0.18 ± 0.003 

to 0.0014 ± 0.0005 d/m/ml. This highest reading was later proved to be a false positive resulting from 

reagent contamination. The results of all subsequent samples were below the detection limit of 0.0004 

d/m/ml based on a sample size of 1000 ml (Horan, 1963). One [role redacted] was involved in more 

than one incident involving [radionuclide redacted]. NIOSH has obtained [radionuclide redacted] 

results for that worker in 1960, 1963, 1964, 1965, 1967, and 1968. Those results are shown in Table 

7-3. 
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Table 7-2: Special Samples Collected at the TRA

Year Urinalysis Plutonium Uranium Transuranic MFP/AP In Vivo

1953 - - - - 338 -

1954 - - - - 668 -

1955 - 3 3 - 1063 -

1956 - - 6 - 1822 -

1957 - - 3 - 3272 -

1958 - - - - 2701 -

1959 - 11 4 1 3215 -

1960 - 4 15 - 3346 -

1961 - 5 1 2 3286 1004

1962 - - 1 - 3139 1890

1963 - 137 1 - 2912 2592

1964 - 25 - 6 2989 1062

1965 - 6 6 1 2297 1086

1966 - 1 2 6 1133 920

1967 - 12 1 - 47 237

1968 - 11 - - 26 310

1969 - 13 - 1 19 261

1970 - - - - 7 460

Note: Values represent special samples collected with the occurrence of alpha-based incidents or where smear survey 

results exceeded TRA guide values. 

Table 7-3 [Radionuclide redacted] Bioassay Results for [Location and Role redacted]

Date Result Units

2/18/1960 <7.00E-04 dpm/mL

3/15/1963 <2.1E-10 uCi/mL

3/16/1964 <4.00E-04 dpm/mL

3/11/1965 <8.00E-04 dpm/mL

3/11/1965 <7.00E-04 dpm/mL

10/6/1967 <4.00E-04 dpm/ml

1/31/1968 <4.00E-04 dpm/mL

4/12/1968 <3.00E-04 dpm/mL
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Airborne Levels

TRA controlled exposure to airborne and re-suspended radioactivity by CAM and smear survey 

programs. Examples of CAM are available (Air Monitoring, 1965-1966, Air Monitoring, Feb1953-

Oct1954, Air Monitoring, Dec1956-Aug1957, Keller, 1958, Air Monitoring, 1966-1967, Air 

Monitoring, 1966, Air Monitoring, Mar1965-Dec1966). Examples of contamination survey data 

available: (Survey, Jun1958, Survey, Nov1958, Survey, Oct1955, Survey, May-Jun1955, Survey, 

Aug-Dec1953, Survey, Apr1963, Survey, Aug1968, Survey, Jul1963, Survey, Apr1969). NIOSH 

does not intend to use air monitoring results to bound potential doses of radiation; rather, these data 

will be used to demonstrate the daily capability of detecting airborne radioactivity.

Alternative Data Sources for Bounding Internal Dose

The contamination control limits employed for the detection and control of released activity beyond 

the control boundaries related to: (1) instrumentation capabilities; and (2) the basic philosophy of the 

acceptance of detectable contamination. The contamination control limits for alpha on plant surfaces, 

and particularly personnel, were set close to the MDA, so that any detectable contamination was a 

signal for preventive and follow-up evaluations and actions. From the onset of operations through the 

end of the 1960s, control levels on smear surveys were set at 500 dpm β and 20 dpm α per 100 cm2

. 

From the 1970s to the termination of CPP operations, smear control levels were set at 300 dpm β and 

20 dpm α per 100 cm2. Table 7-4 provides the total number of smear surveys taken at MTR and ETR 

by year for five of the years for which INL reported the number of smears taken; these data are from 

MTR Health Physics and ETR Health Physics monthly reports that are currently available to NIOSH. 

The smears reported for ETR are lower in 1958 and 1959 as those were early years of operation for 

ETR and MTR. Smear data for TRA facilities for each month are available for data capture at INL.

Table 7-4: MTR Smear Totals

Year No. of MTR 

Surveys Taken

No. of ETR 

Surveys Taken

SRDB Ref ID

1958 81700 15174 87084, 87090

1959 * 12908 87091

1962 * 56450 139240

1965 111400 64200 139164, 139299

1966 69600 74800 139196, 139169

*Only partial smear results available
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7.2.1.2 Methods for Bounding Operational Period Internal Dose for TRA

Uranium

The potential for uranium exposures was minimal because TRA used encapsulated fuel sources. TRA 

Health Physics used air monitoring data and contamination smear data to control potential uranium 

exposure. An interview with a former [organization redacted] staff member stated that exposure to 

airborne uranium was not an issue at TRA (Personal Communication, 2014i). In the few instances in 

which fuel capsules were ruptured in a lab or research setting, TRA collected bioassay samples. 

NIOSH will use these data to reconstruct doses to the exposed workers. 

Fission Products

Workers with potential for exposure to fission and activation products due to releases of radioactive 

materials at TRA reactors, laboratories, and ancillary facilities were routinely monitored by both 

urinalysis and whole-body counting from the onset of radioactive operations at MTR through 1966. 

NIOSH will reconstruct intakes from potential exposures, using both urinalysis and in-vivo monitoring 

records supplied by DOE. Since TRA incorporated an in-vivo counting frequency of twenty-five 

percent of workers per year starting in 1967, NIOSH will develop a co-worker data model for 

fission-product intakes for 1967 through 1970 to complement monitoring data supplied by DOE. 

At TRA, urine samples were typically analyzed only for gross beta, gross gamma, and/or strontium 

radioactivity. For such samples, NIOSH will assess missed Sr-90 and/or Cs-137 intakes in 

accordance with ORAUT-OTIB-0054 and ORAUT-OTIB-0060. Similarly, NIOSH will assess 

missed Cs-137 intakes when using in-vivo data in accordance with ORAUT-OTIB-0060. 

Plutonium and Actinides

A limited known set of workers worked with un-encapsulated plutonium, americium and other 

actinides in TRA hot cells and laboratories, primarily after 1961 in the MTR Alpha Labs. NIOSH will 

reconstruct intakes from known exposures to actinides using worker monitoring data supplied by 

DOE. NIOSH may develop a co-worker actinide model to reconstruct doses for the period 1967 

through 1970 using the monitoring data supplied by DOE for a chemist known to have had high 

potential for actinide exposure (see Table 7-3). 
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7.2.2 Evaluation of Bounding Internal Doses for CPP 

To evaluate the petitioner concerns regarding internal exposures at CPP, NIOSH investigated the 

source term, available personnel monitoring records, and air sample data, as needed. Based on this 

information, NIOSH made a determination regarding the feasibility of reconstructing internal doses. 

The original design philosophy for CPP based the routine monitoring and control program on fission 

products as the controlling radionuclides. That is, if fission products were controlled to permissible 

levels, the alpha radiation hazards would automatically be adequately controlled. However, the 

varying fuels used as feedstock changed this original concept so that, in parts of the process, alpha 

hazards were limiting (Plutonium, 1980).

A tabular representation of the feasibility of reconstructing doses is presented in Table 7-5, in which 

the dark green cells with an “F” indicate reconstructions that are feasible, the light green cells with a 

“C” indicate reconstructions that are feasible but a mixed-fission-product co-worker model is needed, 

and the red cells with an “I” indicate that an infeasibility exists for that particular exposure type. The 

rationale for these feasibility designations is presented in the following subsections.

Table 7-5: Feasibility Summary for CPP (1953-1974)

Exposure 

Source

Years

53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74

MFP F F F F F F F F F F F F F F C C C C C C C C

Pu/Np F F F F F F F F F F I I I I I I I I I I I I

Uranium F F F F F F F F F F I I I I I I I I I I I I

Th/Transuranics F F F F F F F F F F I I I I I I I I I I I I

Other F F F F F F F F F F F F F F C C C C C C C C

Photon F F F F F F F F F F F F F F F F F F F F F F

Neutron F F F F F F F F F F F F F F F F F F F F F F

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible but a mixed-fission-product co-worker model is needed.

I = Dose reconstructions are infeasible.

7.2.2.1 Evaluation of Bounding Process-Related Internal Doses for CPP

Urinalysis and Fecal Information and Available Data 

CPP was designed as a closed process, where dissolution equipment and separation equipment were 

handled remotely (Stroschein, 1967). Process samples and research quantities of radioactive materials 

were handled within glove boxes and remotely handled within the Remote Analytical Facility. There 

were no indications of breaches to the glove boxes in any of the reviewed documents, in any of the 

employee interviews, or in the petition. However, the process-sampling methods and direct-contact 

maintenance structure provided the means by which workers could come in contact with radioactive 

materials. 

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INL had a routine monitoring program for a large portion of its history. That monitoring program was 

based largely on gross-gamma and beta analyses of urine samples. Incident-prompted special analyses 

were also performed, but not to a degree that would allow for the bounding of exposures to 

transuranics considering that the general attitude at CPP toward low-level contamination appeared to 

be too complacent (Rich, 1974). In 1966, the routine collection of urine samples was halted in favor 

of a routine whole-body counting procedure (McCaslin, 1966a). Table 7-6 provides internal 

monitoring records for CPP from 1954 to 1975. 

Table 7-6: CPP-Specific Internal Monitoring Records 

Year β/γ 

Urine

α/β 

Urine

Uranium 

Urine

Plutonium 

Urine

Plutonium 

Fecal

Neptunium 

Fecal

Whole Body 

Counts

1954 403 5 28 - - - -

1955 1029 - 30 - - - -

1956 1166 - 8 - - - -

1957 1729 - 75 1 2 - -

1958 1405 1 98 2 - - -

1959 1298 - 52 10 - - -

1960 645 - 25 1 - - -

1961 729 - - 3 - - 19

1962 807 - 3 18 - - 49

1963 820 - 30 42 - - 33

1964 998 - 27 28 - - 71

1965 884 - 37 35 - - 133

1966 222 - - 8 - - 218

1967 11 - - - - - 99

1968 24 - - - - - 98

1969 19 - - - - - 42

1970 12 1 - 1 - - 100

1971 - - - - - 135

1972 - - 193 86 - - 133

1973 - - - 55 - - 98

1974 1 - 1 29 - 1 229

1975 16 - 2 7 - - 868

Source: Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986

An analysis of the available monitoring data suggests that CPP personnel exposures were chronic in 

nature. Evidence points to the conclusion that shift laboratory personnel were exposed to low levels 

of internal radioactivity for many years, at levels and to isotopes reflective or proportional to the 

levels and isotopes of the work performed during the period under evaluation in this report. 

Exposures recorded in the mid- and late-1970s were detected as a result of a newly-initiated routine 

bioassay program (NIOSH has access to documented evidence that the CPP radiation protection 

program had significantly improved by 1975). The potential for exposures to transuranics that had 

been separated from the mixed fission products makes it unlikely that exposures to alpha-emitters can 

be adequately reconstructed from January 1963 through December 1974. 
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Whole Body Counting Information and Available Data

Groups of individuals who worked continuously with contaminated materials were scheduled for 

whole-body counts by the Health & Safety Services secretary according to the 1963 schedule shown

in Figure 7-1 (Source: McCaslin, 1963).

Figure 7-1: CPP Whole Body Count Frequency Schedule for 1963
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Special analyses would be requested by the Health Physics sections when internal exposure was 

suspected due to events occurring in the plant. The Health and Safety section determined, based on 

job duties, which individuals would receive a termination whole-body count at the time of their 

termination physical examination (McCaslin, 1963).

Personnel exposures for specific functional groups at CPP increased as the plant aged due to increased 

equipment age and the activity of the fuels being processed. This resulted in increased direct 

maintenance. CPP exposures in excess of that normally permitted drew attention from the AEC and 

Allied Chemical Corporation (Rich, 1974). Despite the increasing potential for exposure, the 

monitoring frequency decreased from the 1963 schedule shown in Figure 7-1 to each individual 

receiving a whole-body count every four years (McCaslin, 1966). During this time, bioassay samples 

(fecal and urine) were collected and analyzed only when an exposure incident was suspected. Based 

on this review, it is possible that even a routine whole-body count would be inadequate to detect 

low-level exposures or exposures to alpha- or beta-emitters, much less an incident-based program. 

Within the incident-based program, the typical problems were being encountered in obtaining from 

exposed personnel the cooperation necessary to provide samples as requested, and to furnish all of the 

necessary information (Rich, 1974). 

Increased potential for intake due to poor contamination control and inadequate personnel monitoring 

for exposures to transuranics separated from mixed fission products makes it unlikely that exposures 

to alpha emitters can adequately be reconstructed from January 1963 through December 1974. 

Significant programmatic improvements were identified and initiated in 1974 (ICPP HP Upgrade, 

1971-1980).

NOTE: Based on a need to further evaluate the implementation of the CPP programmatic 

improvements discussed above, NIOSH has reserved the time period beginning January 1, 

1975 for further evaluation. 

Airborne Levels

The Health and Safety Division provided a fixed air monitoring network as part of the complete 

radiation monitoring program (Horan, 1959). A total of 32 continuous air monitors were used in all 

areas of CPP to detect airborne radioactivity. However, continuous monitoring of air activity in the 

cells was difficult to do in a representative manner; as a result, cell entry was made using complete 

respiratory protection until contamination levels were reduced to insignificant levels (Rich, 1974). 

The CPP Health Physics Manual required use of a filter-type respirator when airborne activity 

exceeded 1 x 10-08 μCi/cm3 for beta/gamma activity or 1 x 10-11 μCi/cm3 for alpha activity. An army 

assault-type mask was required when levels exceeded this by a factor of 10; positive-pressure air 

masks were required if there were levels higher by a factor of 1,000 (CPP HP Manual, 1952). There 

was no routine program for air sampling or monitoring in the Mass Spectrometry Lab at the time of 

the 1972 plutonium contamination incident (Malmstrom, 1972).
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Permissible airborne concentration limits were documented in the Process and Laboratory Building 

Operating Manual (Irvine, 1954), as shown in Table 7-7. 

Table 7-7: CPP Permissible Airborne Contamination Levels 

Radionuclide Concentration 

(μCi/cm3 air)

Strontium-89 2 x 10-8

Strontium-90 2 x 10-10

Iodine-131 3 x 10-9

Xenon-133 5 x 10-6

Xenon-135 1.7 x 10-2

Uranium (natural) 1.7 x 10-11

Plutonium-239 2 x 10-12

Fission Products 1 x 10-9

Source: Irvine, 1954

An adequate evaluation of the chronic intake of radioactive materials without extensive bioassay 

samples can be made only by performing a more sensitive and detailed analysis of the filters removed 

from the air sampling devices (Rich, 1974). Beginning in August 1959, CAM filters from the Change 

Room in Building 602, the Access Corridor, the Sample Corridor, the Process Makeup Area, Room 

212, and the Shift Control Lab were saved and counted for long-lived alpha activity (Willis, 1959). 

Although a routine air monitoring program appears to have been in place at CPP, NIOSH has found 

very limited air sampling results. These results include: 

• 569 samples taken in G and H Cells, the Remote Analytical Facility, and Building 602 during 

1955 (G&H Air, 1955). 

• 343 β/γ results and 794 alpha results from Process Makeup Area, Access Corridor, and the Solvent 

Burner during the time period from February 1957 to January 1960 (CPP Air, 1957-1960), 

• 130 α/β results acquired using a Hi-Vol air sampler in the Solvent Burner area, and CAMs in 

Building 605, the Access Corridor, the Makeup Area, and the Operating Corridor during the time 

period from 1960 to 1961 (CPP Hi Vol Air, 1960-1961). 

NIOSH has determined that these limited air monitoring results cannot be used to adequately 

reconstruct exposures to alpha-emitters from January 1963 through December 1974. 

Alternative Data Sources for Bounding Internal Dose

The contamination control limits employed for the detection and control of released activity beyond 

the control boundaries related to: (1) instrumentation capabilities; and (2) the basic philosophy of 

acceptance of detectable contamination. The contamination control limits for alpha on plant surfaces, 

and particularly personnel, were set close to the MDA, so that any detectable contamination was a 

signal for preventive and follow-up evaluations and actions. From the onset of operations through the 
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end of the 1960s, control levels on smear surveys were set at 500 dpm β and 20 dpm α per 100 cm2. 

From the 1970s to the termination of CPP operations, smear control levels were set at 300 dpm β and 

20 dpm α per 100 cm2 (CPP HP Manual, 1952). Table 7-8 provides information on CPP 

contamination monitoring. 

Table 7-8: CPP Contamination Monitoring

Year Surveys Smears Cases Requiring 

Cleaning

SRDB Ref ID

1953 1999 5571 - 138206

1954 11819 24111 - 138206

1955 11697 30623 - 138206

1956 18921 81756 - 138206

1957 13044 53874 637 88592

1958 22879 85661 1596 87078

1959 28699 100021 2382 87080

1960 21631 58089 988 138206

1961 19132 55823 511 138206

1962 16502 65749 453 138206

1963 23262 85082 700 138707

1964 32302 129352 1538 138206

1965 36914 139848 2512 139285

1966 34435 120019 1481 139194

In October 1974, the Preliminary ICPP Health Physics Upgrade Program report was published and 

delivered to management with the intention of identifying and reducing the inherent risks of ICPP 

operations (Rich, 1974). This report provided an overview of the total radiological control program at 

CPP and detailed the interrelationship of each aspect of the program. In the body of the report, an 

attempt was made in each of the listed categories to:

• describe conditions as they existed, with a description of any observed deficiencies 

• list the goals for programmatic improvement, keeping in mind the "State-of-the-Art"

• outline the initial plans to correct any deficiencies identified

• provide a rough estimate of cost for manpower and equipment 

• estimate the time required to complete that phase of the program.

Publication of this report marked the initiation of substantial changes in the radiological control 

program (Rich, 1979).

7.2.2.2 Methods for Bounding Operational Period Internal Dose for CPP

Despite an increasing potential for exposure, monitoring frequency decreased in 1966 so that each 

individual received a whole-body count every four years (McCaslin, 1966). During this time, 

bioassay samples (fecal and urine) were collected and analyzed only when an exposure incident was 

suspected. Based on NIOSH’s review, it is possible that even a routine whole-body count would be 

inadequate to detect low-level exposures or exposures to alpha- or beta-emitters, much less an 
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incident-based program. Within the incident-based program, the typical problems were being 

encountered in getting the cooperation necessary from exposed personnel to obtain samples as 

requested and all necessary information (Rich, 1974). Increased potential for intake due to poor 

contamination control and inadequate personnel monitoring for exposures to transuranics separated 

from mixed fission products makes it unlikely that exposures to alpha emitters can adequately be 

reconstructed from January 1963 through December 1974. 

Uranium

CPP was designed as a closed process, wherein dissolution and separation equipment were handled 

remotely. The first radionuclide-specific bioassay results located for CPP are uranium results. 

NIOSH has determined that it does not have sufficient information to adequately bound exposures to 

uranium from January 1, 1963 through December 31, 1974. 

Neptunium

Prior to 1965, neptunium was disposed of in waste streams. During the neptunium separation 

campaigns that began in 1965, the potential for exposure to this element increased for workers who

drew samples from the process streams, as well as for analytical staff, who measured the neptunium 

concentration in the storage tanks and process streams. Although low, there was a potential for 

workers to be exposed to neptunium during activities at the CPP. There was no routine monitoring 

program for neptunium intakes and it is likely that the incident-based monitoring program could have 

missed chronic, low-level exposures. NIOSH has determined that it does not have sufficient 

information to adequately bound exposures to neptunium from January 1, 1963 through December 31, 

1974. 

Plutonium

Prior to 1965, plutonium was considered an undesirable contaminant of the uranium product and was 

disposed of in waste streams. During the neptunium separation campaigns that began in 1965, the 

potential for exposure to plutonium increased for workers who drew samples from the process 

streams, as well as for analytical staff, who measured the plutonium concentration in the storage tanks 

and process streams. Although low, there was a potential for workers to be exposed to plutonium 

during activities at CPP. There was no routine monitoring program for plutonium intakes and it is 

notable that the incident-based monitoring program is known to have missed chronic, low-level 

exposures, as described in Section 5.4.2. NIOSH has determined that it does not have sufficient 

information to adequately bound exposures to plutonium from January 1, 1963 through December 31, 

1974. 

Fission Products

From the beginning of operations, bioassay results for fission products were used as an indicator of 

contamination control. For urine samples only analyzed for gross beta, gross gamma, and/or 

strontium radioactivity, NIOSH will assess missed Sr-90 and/or Cs-137 intakes in accordance with 

ORAUT-OTIB-0054 and ORAUT-OTIB-0060. Similarly, NIOSH will assess missed Cs-137 intakes 

when using in-vivo data in accordance with ORAUT-OTIB-0060. Based on the procedural 

information and the data on hand, NIOSH finds that it has adequate monitoring data to allow for 
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sufficiently accurate estimation of internal fission product doses for workers during the period from 

January 1, 1954 through December 31, 1974. 

Barium/Lanthanum

The RaLa process campaign ran from 1957 to 1963, during which time, over a million curies of 

Ba-140 were recovered. The fuel elements used in the RaLa process were intensely radioactive 

because of the short time that had elapsed between removal from the reactor and processing. 

Recovery of La-140 was a complicated process completed in a specialized processing cell (L-Cell) in 

CPP-601. Safety measures included: isolation of the process behind a shielded viewing window, 

remote-handling manipulators, a periscope, and an intercom system installed in the vacant L-Cell to 

allow an operator to see and hear the work going on behind six feet of concrete shielding. The nature 

of the process makes it extremely unlikely that a worker would be exposed to these materials. NIOSH 

has found no indication that there were ruptures or leaks in the process shielding or equipment, and no 

indication of barium or lanthanum contamination outside the cell.

During RaLa runs, the CAM in the PM area near L-Cell was to be checked for possible airborne 

contamination. In addition, once each hour, an individual designated by the HP foreman was assigned 

to routinely check the following: (1) all RaLa piping in the PM-PO-PA areas for any change in 

radiation; and (2) all CAMs in CPP-601, CPP-604, and CPP-643 for any rise in air activity. 

The intense radiation of the product (approximately 60,000 rem/hr at one meter) necessitated remote 

handling of product materials. At the conclusion of RaLa processing, the RaLa product was remotely 

placed in the shipping cask inside the cell. A physicist was required to be present when the L-Cell 

hatch cover was removed, the shipping cask was removed from the cell, and the cask prepared for 

shipment. Full-face masks equipped with carbon filters were required during this part of the 

operation, and after the RaLa shipping container left the PM area, the physicist took smear samples of 

the PM area floor in the vicinity of L-Cell.

The air from the cell was required to be monitored continuously whenever the cell top was open and 

during decontamination. Any cell entry prior to or during decontamination required airline respirators 

or full-face air-purifying masks. After decontamination was complete and the air levels were 

measured to be less than 10 RCG potential, carbon respirators with static web filters were permitted 

for cell entry (HP Regs, 1963). 

Lanthanum-140 decays by beta emission and, in doing so, emits several photons that should be 

detectable via whole-body counting. In addition, the presence of La-140 would have been detected 

during a urinalysis that was analyzed for gross gamma or beta counting. Based on the process 

information and the data on hand, NIOSH finds that it has adequate monitoring data to allow for 

sufficiently accurate estimation of barium-140 or lanthanum-140 exposures for workers during the 

period from January 1, 1954 through December 31, 1974. 
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Thorium

Two thorium slugs were dissolved in 1957. Each of these had a total mass of 1730 g, with 78 grams 

being aluminum cladding. Separation of the uranium was performed in a glove box and small 

samples were transferred to the Mass Lab for analysis, minimizing the exposure potential.

Six slugs were dissolved in 1961. Each slug had a mass of about 1730 g; they were dissolved and 

prepared for analysis in the Multicurie Cell. Manipulations within the cell were made with a pair of 

master-slave manipulators, minimizing the chance for an intake. 

With a specific activity of 1.1 x 10-7 Ci/g, there is no apparent plausible scenario that would have 

allowed for the ingestion or inhalation of an appreciable amount during either of the above campaigns. 

Therefore, additional analysis of thorium, beyond the bounding approaches for other radionuclides 

that has been performed for this evaluation, has not been performed. Specific analyses for thorium 

will be addressed on a case-by-case basis if a plausible exposure scenario is identified or presented.

Rare Gases

Xenon and krypton are readily-diffusible gases that are neither used nor produced by the body. Most 

of the xenon or krypton gas that enters the circulation from a single breath is returned to the lungs and 

exhaled after a single pass through the peripheral circulation; they are essentially inert. Radiation 

doses from inhaling or ingesting krypton or xenon are small compared to the dose from external 

radiation, such as could occur in a cloud of gas. Thus, internal exposures due to these gases will not 

be considered further in this report. 

Iodine

The RaLa process ran from February 1957 to August 1963. Because of a cooling time of only two 

days (compared to the 90-120 days cooling normally provided for most other fuel elements before 

reprocessing), the RaLa process contributed large quantities of the fission product iodine to the 

process off-gas. The virtually-complete containment of iodine was a major objective in the design of 

the RaLa process. However, the first processing of a fully-irradiated fuel element disclosed serious 

deficiencies in the iodine-containment measures. 

An upgraded RaLa Off-Gas system contained charcoal beds for removing radioiodine and a gas holder 

for storing the noble gases during decay (Waste Management, 1974). This system was able to remove 

gaseous iodine with a 99.9% efficiency, but was still limited in its ability to remove iodine-bearing 

dusts and mists (I-131 Containment, 1961). Site-wide routine samples were initially analyzed only for 

gross alpha and beta, with isotopic analysis performed only if activity was discovered in the sample - 

although capabilities existed for analyses for specific isotopes such as Pu-239, Sr-90, Co-60, and I131 in body excreta, water, and air samples. Beginning in May 1957, the Division started requesting 

gross-gamma counts on all urine samples submitted by CPP personnel in an effort to detect the 

potential uptake of iodine from fresh fuel (HP Monthly Reports, 1957). NIOSH intends to use these

urinalysis data to reconstruct doses to the exposed workers. 
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Tritium

Historical records on INL tritium exposures are sparse, although it appears that limited environmental 

monitoring began in 1960 (Horan, 1961). The most important tritium source term appears to be HTO 

as a by-product waste material from processing alloy fuels at CPP. Sampling and analysis were 

consistent with methods used at other AEC facilities. In 1974, INL reported a detection limit of 0.6 

kBq/L for tritium in urine using liquid-scintillation counting methods with a 10-minute counting time. 

The year 1974 is also the first year for which NIOSH has tritium-specific bioassay. This does not 

mean that tritium bioassays were not conducted, however. Documentation points to urine samples 

being analyzed for tritium as early as 1968 (Puphal, 1994). However, the results of in-vivo and 

in-vitro bioassay were not included in an individual’s file unless a dose assessment was made from a 

positive result (Tiger Team, 1991).

As described in Section 5.2.2, tritium was likely only present where mixed fission and activation 

products were present. In addition, the source term for tritium and its ability to contribute to the 

workers’ internal doses was likely insignificant compared to the workers’ exposures to mixed fission 

and activation products because the process equipment and enclosed cell structure limited the 

potential for tritium exposure. Based on this information NIOSH feels it can adequately bound tritium 

exposures using the presence of mixed fission products as an indicator of the presence of tritium. 

7.2.3 Evaluation of Bounding Internal Doses for TAN 

The philosophy of TAN’s internal dose monitoring program established a routine monitoring and 

control program based on mixed fission products as the controlling radionuclides. That is, if the 

mixed fission products were controlled to permissible levels, the alpha hazards would be adequately 

controlled. However, potential exceptions to this were noted in Section 5.2.3. 

TAN workers participated in bioassay programs for potential intakes of mixed fission products by 

either urinalysis or whole-body counts, either routinely or when workplace indicators (e.g., elevated 

airborne radioactivity, contamination event, positive nasal smear) indicated that an intake may have 

occurred. When workplace indicators indicated that an intake may have occurred, “special” (nonroutine) bioassay would be requested by TAN Health Physics.

As indicated in Section 5.2.3, the categories of radionuclides at TAN included: mixed fission and 

activation products, radioiodines, radioactive noble gases, plutonium, neptunium, uranium, other 

actinides, and other radionuclides. 

A tabular representation of the feasibility of reconstructing doses is presented in Table 7-9, in which 

the dark green cells with an “F” indicate reconstructions that are feasible, the light green cells with a 

“C” indicate reconstructions that are feasible but a mixed-fission-product co-worker model is needed, 

and the yellow cells with an “R” indicate that years for that particular exposure type are reserved for 

further evaluation. The rationale for these feasibility designations is presented in the following 

subsections. 
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Table 7-9: Feasibility Summary for TAN (1955-1970)

Exposure Source

Years

55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70

Mixed Fission & Activation Products F F F F F F F F F F F F C C C C

Radioiodines F F F F F F F F F F F F C C C C

Actinides with MFP Present F F F F F F F F F F F F C C C C

Uranium without MFP Present F F F F F F R R R R R R R R R R

Other Radionuclides F F F F F F F F F F F F C C C C

External Photon and Electron F F F F F F F F F F F F F F F F

External Neutron F F F F F F F F F F F F F F F F

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible but a mixed fission product co-worker model is needed.

R = Reserved for further evaluation.

7.2.3.1 Evaluation of Bounding Process-Related Internal Doses for TAN

Urinalysis Information and Available Data 

From September 1955 through January 1970, approximately 9,467 urinalyses for gross beta 

radioactivity, strontium radioactivity, and/or gross gamma radioactivity were performed for TAN 

workers in order to detect intakes of mixed fission and activation products. The urinalyses for gross 

gamma radioactivity could also detect intakes of radioiodines and other gamma-emitting 

radionuclides. The urinalyses for beta-gamma radioactivity could also detect intakes of other 

non-volatile beta-emitting radionuclides.

During 1956, 1962–1966, and 1970, approximately 111 urine samples were specifically analyzed for 

uranium. In May 1958, six urine samples were specifically analyzed for iodine-131. On October 16, 

1962, the urine samples of several TAN workers were analyzed for thorium intakes. Only two urine 

samples from the same TAN worker were specifically analyzed for plutonium, one in 1963 and one in 

1964. On November 13, 1956, one urine sample was analyzed for Po-210. 

No urinalysis results were found for the other radionuclides at TAN. 

Whole Body Counting Information and Available Data

Based on NIOSH’s review of the available data, from July 1961 through 1975, whole-body counts 

were performed for TAN workers. Furthermore, during the period from July 1961 through January 

1970, whole-body counts transitioned into being the primary bioassay method for detecting intakes of 

fission and activation products in TAN workers. Whole-body counts could also detect intakes of 

radioiodines and other gamma-emitting radionuclides. The TAN Whole Body Count Frequency 

Schedule from 1963 indicates that 22.3% of the TAN workers received whole-body counts every six 

months, 21.6% received annual whole-body counts, and 56.1% did not receive any whole-body counts 

(McCaslin, 1963). Based on the worker categories provided in that schedule, the frequencies were 

based on the workers’ potential to receive internal dose (McCaslin, 1963). In addition, all 
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terminations requiring a physical examination, as determined by the Health & Safety sections, also 

required a whole-body count (McCaslin, 1963). 

Airborne Levels

A limited amount of air monitoring data was captured for the TAN facilities for the years 1962-1970 

(CAM 1967-1970, CAM, 1962-1963, CAM, 1963-1964, CAM, 1964-1967).

Alternative Data Sources for Bounding Internal Dose

A limited amount of contamination survey data was captured for the TAN facilities for the years 

1964–1968 (TAN Monitoring, 1963).

7.2.3.2 Methods for Bounding Operational Period Internal Dose for TAN

Mixed Fission and Activation Products

The potential intakes of mixed fission and activation products for monitored and unmonitored TAN 

workers can be estimated using the approaches described in ORAUT-TKBS-0007-5. For urine 

samples analyzed for gross beta, gross gamma, and/or strontium radioactivity, NIOSH will assess 

Sr-90 and/or Cs-137 intakes in accordance with ORAUT-TKBS-0007-5 and ORAUT-OTIB-0060. 

For whole body counts, NIOSH will assess Cs-137 intakes in accordance with ORAUT-TKBS-0007-5 

and ORAUT-OTIB-0060. In accordance with the guidance in ORAUT-OTIB-0054, NIOSH will use 

Sr-90 and/or Cs-137 intakes to estimate the intakes of the other mixed fission and activation products 

that may have been present, unless a bioassay result was available for a specific radionuclide. In the 

event that a bioassay result was available for a specific radionuclide, NIOSH will use that result in lieu 

of the ORAUT-OTIB-0054 approach. 

Based on the procedural information and the data on-hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to mixed 

fission and activation products for TAN workers during the period from 1955 through 1970. 

Radioiodines

The potential intakes of radioiodines without the other mixed fission products present can be 

estimated for monitored and unmonitored TAN workers using the approaches described in ORAUTTKBS-0007-5. Urine samples analyzed for gross gamma radioactivity and whole-body counts can be 

used to assess any potential intakes of radioiodines without the other mixed fission products present. 

Based on the availability of the gross gamma urine sample data and whole-body count data for TAN 

workers, NIOSH finds that it has adequate monitoring data to allow for sufficiently accurate 

estimation of internal doses attributable to radioiodines for TAN workers during the period from 

August 1958 through 1970. 
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Radioactive Noble Gases

Because radioactive noble gases do not represent a significant internal exposure concern, NIOSH 

finds there is no need to assess the internal doses attributable to radioactive noble gases.

Actinides with Mixed Fission Products Present

The majority of the actinides at TAN were present with mixed fission products. The potential intakes 

of actinides when mixed fission products were present can be estimated using the approach described 

in Section 5.5.2 of ORAUT-TKBS-0007-5. Based on that approach, actinide intakes are estimated by 

multiplying the assessed Sr-90 and/or Cs-137 intakes by actinide-to-Sr-90 and/or actinide-to-Cs-137 

ratios.

Based on the procedural information and the data on-hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to actinides 

when mixed fission products were present for TAN workers during the period from 1955 through 

1970. 

Uranium without Mixed Fission Products Present

NOTE: During research for the SEC-00219 evaluation report, it was discovered that uranium 

contamination without mixed fission products was present in the Fuel Storage Vaults of the 

TAN-607 Building and the TAN-615 Building. Further investigation is needed to determine 

the source of the contaminants, the beginning and end dates for the presence of the uranium 

contamination, and whether the exposures to that uranium can be bounded. 

Other Radionuclides

Even though none of the bioassay methods at TAN could detect H-3 and/or C-14, these two 

radionuclides were likely only present where mixed fission and activation products were present. 

Furthermore, the source terms for H-3 and C-14 and their ability to contribute to the workers’ internal 

doses were likely insignificant compared to the workers’ exposures to mixed fission and activation 

products. 

The potential intakes of Na-24 can be estimated for monitored and unmonitored TAN workers using 

the guidance provided in Section 5.5.8 of ORAUT-TKBS-0007-5. Urine samples analyzed for gross 

gamma radioactivity and whole-body counts can be used to assess any potential intakes of Na-24. 

Based on the availability of the gross beta urine sample data, gross gamma urine sample data, and 

whole-body count data for TAN workers, NIOSH finds that it has adequate monitoring data to allow 

for sufficiently accurate estimation of internal doses attributable to Na-24 for TAN workers during the 

period from 1955 through 1970. 
 

 

 

 

 

 

 


 


 


 

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7.2.4 Evaluation of Bounding Internal Doses for Misc. Reactor Areas 

The Miscellaneous Reactor Areas’ internal dose monitoring program established a routine monitoring 

and contamination control program based on mixed fission products as the controlling radionuclides. 

Workers participated in bioassay programs for potential intakes of mixed fission products by either 

urinalysis and/or whole-body counts. Participation was on a routine basis or as required by 

radiological indicators. When workplace indicators indicated that an intake may have occurred, 

“special” (non-routine) bioassay would be requested by the area Health Physics staff. 

A tabular representation of the feasibility of reconstructing doses is presented in Table 7-10, in which 

the dark green cells with an “F” indicate reconstructions that are feasible, the light green cells with a 

“C” indicate reconstructions that are feasible but a mixed-fission-product co-worker model is needed, 

and the yellow cells with an “R” indicate that years for that particular exposure type are reserved for 

further evaluation. The rationale for these feasibility designations is presented in the following 

subsections. 

Table 7-10: Feasibility Summary for Misc. Reactor Areas (1955-1970) 

Exposure Source 

Years 

55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 

Mixed Fission Products F F F F F F F F F F F F C C C C 

Plutonium/Neptunium F F F F F F F F F F F F C C C C 

Uranium F F F F F F F F F F F F C C C C 

Other F F F F F F F F F F F F C R C C 

Photon F F F F F F F F F F F F F F F F 

Neutron F F F F F F F F F F F F F F F F 

F = Dose reconstructions are feasible.
 

C = Dose reconstructions are feasible but a mixed fission product co-worker model is needed.
 

R = Reserved for further evaluation.
 

7.2.4.1 Evaluation of Bounding Process-Related Internal Doses for Misc. Reactor Areas 

Urinalysis Information and Available Data 

The Misc. Reactor Areas was a “proof of principle” area where various reactor projects were built and 

operated. Potential internal exposure to radioactive materials from operation and maintenance 

activities was recognized and monitored. The principal source of internal radiation doses for members 

of the class working in the Misc. Reactor Facilities was mixed fission products, although there was 

limited exposure potential to some exotic radionuclides, noble gases, iodine, and uranium, as 

discussed in Section 5.2.4. 

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The Misc. Reactor Area routine urinalysis program was based on gross beta and gamma analyses of 

urine samples. Positive measurements of interest would initiate identification of the contaminant of 

interest. Incident-driven or special urinalysis was also performed if an internal exposure was 

suspected from radiological conditions (Special HP Services, 1968; Personal Communication 2014i). 

Urinalysis data are available for the Misc. Reactor Areas during the years of operations, except for 

1967-1970. The lack of in-vitro bioassay data between these years was due to transition to in-vivo

measurements with subsequent decreases in in-vitro bioassay (McCaslin, 1966). Table 7-11 provides 

internal monitoring results for the Misc. Reactor Areas.

Table 7-11: Miscellaneous Reactor Area-Specific Internal Monitoring Results

Year

SPERT OMRE ARA

β/γ

urinalysis In-vivo β/γ urinalysis In-vivo β/γ urinalysis In-vivo

1954 - - - - - -

1955 3 - - - - -

1956 80 - - - - -

1957 124 - 69 - - -

1958 122 - 169 - - -

1959 76 - 99 - 81 -

1960 64 - 140 - 90 -

1961 150 4 43 22 413 48

1962 108 9 308 5 12 55

1963 48 27 108 75 52 63

1964 167 48 - - 9 12

1965 142 68 - - - 2

1966 130 16 - - - -

1967 - 19 - - - -

1968 - 14 - - - 1

1969 - - - - - -

1970 - 17 - - - 11

Source: Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986; WBC1, 1961-1977; WBC2, 

1963-1995; WBC3, 1981-1994; WBC4, 1979-1995; WBC5, 1973-1996
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Whole Body Counting Information and Available Data

Table 7-11 presents both in-vivo and in-vitro data for the Misc. Reactor Areas. As with in-vitro

bioassay, event-driven in-vivo measurements would be performed if deemed necessary by Health 

Physics. Of the Misc. Reactor Areas, only SPERT had a counting frequency prescribed by Health and 

Safety, shown in Table 7-12 (McCaslin, 1963). 

Table 7-12: SPERT Whole Body Count Monitoring Frequency 

Group Semiannual Annual Not Required

HP Technicians 4 1 -

SPERT I 4 4 -

SPERT II 5 7 -

SPERT III 5 1 -

SPERT IV 5 4 -

General Area - 8 54

Source: McCaslin, 1963

Airborne and Contamination Levels

Air monitoring and contamination smear programs were used in the Misc. Reactor Areas to control 

personnel exposures to airborne radioactivity and surface contamination. Contamination smear 

programs were well documented in monthly Health Physics reports (SPERT HP Reports, 1959). 

Continuous air monitors were used in all reactor facilities. The SPERT reactors used a single CAM 

on the stack exhaust to measure air effluents which were reported in monthly radioactive waste 

reports. Airborne concentrations tended to be in the μCi-to-mCi range except during transient testing. 

During months when transient tests were performed, airborne releases were much higher but the 

composition of the releases was of fresh fission products with short half-lives. In addition, controlled 

access and favorable meteorological conditions were required for testing (SABER, 1990). Examples 

of the use of CAMs in the ARA, OMRE, and SPERT reactors are available (SL-1 Safety, 4Q1959; 

OMRE HP Survey, 1962-1963; SPERT, 1957, respectively). NIOSH does not intend to use air 

monitoring and contamination results to bound potential internal doses. 

7.2.4.2 Methods for Bounding Operational Period Internal Dose for Misc. Reactor Areas

Fission Products

For the Misc. Reactor Areas, mixed fission and activation products have always been present in 

contaminants. For urine samples only analyzed for gross beta, gross gamma, and/or strontium 

radioactivity, NIOSH will assess missed Sr-90 and/or Cs-137 intakes in accordance with ORAUTOTIB-0054 and ORAUT-OTIB-0060. Similarly, NIOSH will assess missed Cs-137 intakes when 

using in-vivo data in accordance with ORAUT-OTIB-0060. Based on the procedural information and 

the data on-hand, NIOSH finds that it has adequate monitoring data to allow for sufficiently accurate 

estimation of internal fission product doses for workers during the period from January 1, 1955 

through December 31, 1970. 
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Other Radionuclides

As discussed, the primary exposure source at the Misc. Reactor Areas was fission products, and any 

other radionuclides were present with mixed fission products. The potential intakes of other 

radionuclides when mixed fission products were present (as indicated by data in personnel records) 

can be estimated on a case-by-case basis using the approach described in ORAUT-TKBS-0007-5. 

Based on the procedural information and the data on-hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to other 

radionuclides for Misc. Reactor Area workers during the period from January 1, 1955 through 

December 31, 1970. 

NOTE: In 1968, the ARA hot cell was extensively modified for work separating Pa-233, a highspecific-activity alpha-emitter. Modifications included upgrades to the ventilation system and 

addition of isolation boxes for the Pa-233. ARA-I is being reserved for 1968 for further 

evaluation of the separation of Pa-233 that occurred in the hot cells.

7.2.5 Evaluation of Bounding Internal Doses for CFA 

The Central Facilities Area’s internal dose monitoring program established a routine monitoring and 

contamination control program based on mixed fission products as the controlling radionuclides. 

Workers participated in bioassay programs for potential intakes of mixed fission products by either 

urinalysis and/or whole-body counts. Participation was on a routine basis or as required by 

radiological indicators. When workplace indicators indicated that an intake may have occurred, 

“special” (non-routine) bioassay would be requested by the area Health Physics staff.

A tabular representation of the feasibility of reconstructing doses is presented in Table 7-13, in which 

the dark green cells with an “F” indicate reconstructions that are feasible, and the light green cells 

with a “C” indicate reconstructions that are feasible but a mixed-fission-product co-worker model is 

needed. The rationale for these feasibility designations is presented in the following subsections. 

Table 7-13: Feasibility Summary for Central Facilities Area (1949-1970)

Exposure 

Source

Years

49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70

Mixed 

Fission 

Products

F F F F F F F F F F F F F F F F F F C C C C 

Plutonium/

Neptunium F F F F F F F F F F F F F F F F F F C C C C 

Uranium F F F F F F F F F F F F F F F F F F C C C C 

Other F F F F F F F F F F F F F F F F F F C C C C 

Photon F F F F F F F F F F F F F F F F F F F F F F 

Neutron F F F F F F F F F F F F F F F F F F F F F F 

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible but a mixed fission product co-worker model is needed.
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7.2.5.1 Evaluation of Bounding Process-Related Internal Doses for CFA

Urinalysis Information and Available Data 

The Central Facilities Area provided support services to the rest of INL. The primary internal 

exposure concerns would have been from the CF-699 Laundry facility and transportation support for 

radioactive waste shipments. The principal source of internal radiation doses for members of the class 

working in the CFA was mixed fission and activation products, although there was limited exposure 

potential to uranium, as discussed in Section 5.2.5. 

The Health Services Laboratory was located in the CFA. The laboratory was part of AEC Health and 

Safety Technical Services and provided radiochemical analyses of body excreta samples and in-vivo

measurements for gamma-emitting radionuclides. Also located in CFA was the CF-603 medical 

dispensary where in-vitro samples could be delivered (due to its central location). Baseline, 

termination, and excreta samples were often submitted at CF-603. Facility codes for CFA were 

frequently used for in-vitro samples and in-vivo measurements although the worker was not assigned 

to CFA. Consequently, a disproportionately high number of results are attributed to CFA. The data 

are reported in Table 7-14. 

Table 7-14: Central Facilities Area-Specific Internal Monitoring Results 

Year β/γ

urinalysis

U 

urinalysis

Sr 

urinalysis

I 

urinalysis

Pu 

urinalysis 

and fecal

In-Vivo 

1953 4 1 1 - - -

1954 196 7 - -

1955 410 2 - - - -

1956 386 - - - - -

1957 632 - - 29 2 -

1958 303 12 - 10 - -

1959 953 - - 10 - -

1960 1225 - 17 9 - -

1961 1679 1 433 - - 110

1962 1277 - 793 - - 284

1963 1146 - 360 - 5 356

1964 1177 - 804 - 9 364

1965 1299 1 625 - 2 452

1966 330 - 432 - - 313

1967 111 10 263 - - 80

1968 - - - - - 101

1969 - - - - - 84

1970 55 3 - - - 105

Source: Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986; WBC1, 1961-1977; WBC2, 

1963-1995; WBC3, 1981-1994; WBC4, 1979-1995; WBC5, 1973-1996
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Whole Body Counting Information and Available Data

Table 7-14 presents in-vivo and in-vitro bioassay data for the CFA. As with in-vitro bioassay, 

event-driven in-vivo measurements would be performed if deemed necessary by Health Physics. The 

CFA was not on a prescribed whole-body counting schedule like TRA, CPP, TAN and SPERT 

(McCaslin, 1963). 

Airborne and Contamination Levels

Air monitoring and contamination smear programs were used in the CFA to control personnel 

exposures to airborne radioactivity and surface contamination. Contamination controls in the CF-699 

Laundry included radiation monitors, a monitoring table for laundered clothing, and a CAM. The 

radiation monitor was located above the laundry receiving door and set to alarm at 2 mR/hr (Savignac, 

1961). The Laundry monitoring table employed eight GM tubes and two scaling units that allowed it 

to distinguish between contamination spots vs. distributed contamination on clothing. A CFA health 

physicist routinely checked the CF-699 monitoring equipment and was responsible for responding to 

alarms. The CFA cafeteria was also routinely surveyed and smeared (CFA HP Logs, 1965-1966). 

CFA Health Physics monthly reports provide summary data on the number of smears taken, number 

of special air samples, and the number of vehicle checks for contamination. Also provided are the 

numbers of on-site and off-site radioactive shipments (Savignac, 1961; Johnson, 1961). Analyses of 

sludge from the CF-674 Sewage Treatment Plant, to which the CF-699 Laundry facility was the 

primary contributor, identified mixed fixed products, such as Co-60, Zr-95, Nb-95, Cs-137, and Ce144 (Savignac, 1961). NIOSH does not intend to use air monitoring or contamination results to bound 

potential internal doses.

7.2.5.2 Methods for Bounding Operational Period Internal Dose for CFA

Fission Products

For the CFA, mixed fission and activation products were the internal dose hazards of concern. For 

urine samples only analyzed for gross beta, gross gamma, and/or strontium radioactivity, NIOSH will 

assess missed Sr-90 and/or Cs-137 intakes in accordance with ORAUT-OTIB-0054 and ORAUTOTIB-0060. Similarly, NIOSH will assess missed Cs-137 intakes when using in-vivo data in 

accordance with ORAUT-OTIB-0060. Based on the procedural information and the data on-hand, 

NIOSH finds that it has adequate monitoring data to allow for sufficiently accurate estimation of 

internal fission product doses for workers during the period from January 1, 1953 through December 

31, 1970. Prior to 1953, NIOSH will assess internal dose in accordance with ORAUT-OTIB-0033. 
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Other Radionuclides 

As discussed, the primary exposure source at the CFA was fission products and any other 

radionuclides present with mixed fission products. The potential intakes of other radionuclides when 

mixed fission products were present (as indicated by data in personnel records) can be estimated on a 

case-by-case basis using the approach described in ORAUT-TKBS-0007-5. 

Based on the procedural information and the data on-hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to other 

radionuclides for CFA workers during the period from January 1, 1953 through December 31, 1970. 

7.2.6 Evaluation of Bounding Internal Doses for the Burial Ground 

The Burial Ground’s internal dose monitoring program was based on a strict contamination control 

program with entry and exit monitoring. With the exception of Rocky Flats waste, mixed fission 

products were considered the controlling radionuclides. When workplace indicators indicated that an 

intake may have occurred, “special” (non-routine) bioassay would be requested by the area Health 

Physics staff.

A tabular representation of the feasibility of reconstructing doses is presented in Table 7-15, in which 

the dark green cells with an “F” indicate reconstructions that are feasible, the light green cells with a 

“C” indicate reconstructions that are feasible but a mixed-fission-product co-worker model is needed, 

and the yellow cells with an “R” indicate that years for that particular exposure type are reserved for 

further evaluation. The rationale for these feasibility designations is presented in the following 

subsections.

Table 7-15: Feasibility Summary for the Burial Ground (1952-1970)

Exposure Source

Years

52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70

Mixed Fission 

Products F F F F F F F F F F F F F F F C C R R 

Plutonium/Neptunium F F F F F F F F F F F F F F F C C R R 

Uranium F F F F F F F F F F F F F F F C C R R 

Other F F F F F F F F F F F F F F F C C R R 

Photon F F F F F F F F F F F F F F F F F F F 

Neutron F F F F F F F F F F F F F F F F F F F 

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible but a mixed fission product co-worker model is needed.

R = Reserved for further evaluation.

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7.2.6.1 Evaluation of Bounding Process-Related Internal Doses for the Burial Ground

Urinalysis and Whole Body Counting Information and Available Data 

The Burial Ground served as a solid radioactive waste disposal area for INL-generated solid waste, an 

interim waste burial facility for the western United States (Hayden, 1961), and Rocky Flats Plant 

waste beginning in 1954. The principal source of internal radiation doses for members of the class 

working at the Burial Ground was mixed fission and activation products, although there was exposure 

potential to plutonium from Rocky Flats waste, as discussed in Section 5.2.6. Any potential exposure 

to exotic radionuclides would be dominated by mixed fission product exposures at the Burial Ground. 

Responsibility for the Burial Ground during the period of this evaluation resided with the CFA Health 

Physicist. A pool of workers known as “yardmen” comprised the work force that emplaced waste at 

the Burial Ground and worked out of CFA during those work evolutions. The in-vitro and in-vivo

measurements for monitoring at the Burial Ground are included in Table 7-14 in Section 7.2.5.1. 

Review of worker names on SWPs indicate that, at a minimum, some workers were monitored by 

bioassay, although it is possible that the monitoring was due to work activities in other areas. Special 

bioassay also exists, but the results could not be directly related to a contamination event at the Burial 

Ground. Requests for incident or non-routine bioassay would be made by the CFA Health Physicist. 

Airborne and Contamination Levels

Responsibility for the overall operation of the Burial Ground resided with the CFA Health Physicist 

(Hayden, 1962). These responsibilities also included preparation of monthly reports, which included 

the number of off-site and on-site shipments by area and curie content. Examples are available 

(Hueter, 1963; Savignac, 1963; Sauvignac, 1963a; Hazards Control Qtr, 1965). Waste disposal at the 

Burial Ground required the presence of a health physicist. A review of Burial Ground SWPs from 

January 1962 to December 1964 indicate the level of monitoring that was required for certain work 

evolutions. Examples include personal survey upon completion of work, hand and shoe counts, and 

the level (continuous or intermittent) of health physics coverage (SWP, 1962-1964).

Table 7-16 presents the waste burial schedule for Rocky Flats Plant waste from February through 

August 1963. It is representative of the waste burial practices for Rocky Flats waste sent to INL for 

burial. Of note is that burial of Rocky Flats waste was performed only for a limited number of days 

each month. 
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Table 7-16: Example of Rocky Flats Waste Burial Schedule, February-August 1963 

Burial Summary Report Ft3 Buried No. of Pieces Date Buried

February 1963 3542 a481 February 1, 1963

February 1963 2352 320 February 15, 1963

March 1963 3289 405 March 1, 1963

March 1963 3395 462 March 15, 1963

March 1963 3426 466 March 22, 1963

March 1963 3022 340 March 29, 1963

April 1963 3881 46 April 19,1963

April 1963 4121 43 April 26, 1963

May 1963 2272 309 May 6, 1963

May 1963 2411 328 May 10, 1963

May 1963 2749 374 May 17, 1963

May 1963 3293 448 May 24, 1963

June 1963 3292 448 June 7, 1963

June 1963 3277 448 June 14, 1963

June 1963 4446 238 June 21, 1963

June 1963 4332 48 June 28, 1963

July 1963 4270 55 July 5, 1963

July 1963 3161 430 July 12, 1963

July 1963 2866 390 July 19, 1963

August 1963 3287 449 August 9, 1963

August 1963 2301 313 August 16, 1963

August 1963 3241 441 August 23, 1963

August 1963 4466 549 August 30, 1963

Source: Waste Shipments, Feb1963; Waste Shipments, Mar1963; Waste Shipments, Apr1963; Waste Shipments, 

May1963; Waste Shipments, Jun1963; Waste Shipments, Jul1963; Waste Shipments, Aug1963; Waste Shipments, 

Nov1963

a Estimated – record illegible

7.2.6.2 Methods for Bounding Operational Period Internal Dose for the Burial Ground

NOTE: In late 1969, some experimental equipment that was accidentally buried in Pit 1 was 

retrieved; this marked the first time in the history of the Burial Ground that buried waste was 

deliberately exhumed. Further NIOSH evaluation is required of the personnel monitoring 

data for waste-retrieval activities, necessitating reservation of the Burial Ground from 1969 

forward. 

Mixed Fission and Activation Products

For the Burial Ground, mixed fission and activation products were the primary internal dose hazards 

of concern. For urine samples only analyzed for gross beta, gross gamma, and/or strontium 

radioactivity, NIOSH will assess missed Sr-90 and/or Cs-137 intakes in accordance with ORAUTOTIB-0054 and ORAUT-OTIB-0060. Similarly, NIOSH will assess missed Cs-137 intakes when 

using in-vivo data in accordance with ORAUT-OTIB-0060. Based on the procedural information and 

the data on-hand, NIOSH finds that it has adequate monitoring data to allow for sufficiently accurate 

estimation of internal fission product doses for workers during the period from January 1, 1953 
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through December 31, 1968. Prior to 1953, NIOSH will assess internal dose in accordance with 

ORAUT-OTIB-0033. 

Plutonium

The radiological monitoring program at the Burial Ground included the presence of a health physicist, 

safe work permits for all waste disposals, personnel surveys upon completion of work, air monitoring, 

and decontamination of vehicles at CPP if they were found to be contaminated. This defense-in-depth 

approach was adequate to ensure that unmonitored intakes of plutonium did not occur. 

As discussed, the primary exposure source at the Burial Ground was fission products and any other 

radionuclides present with mixed fission products. The potential intakes of other radionuclides when 

mixed fission products were present (as indicated by data in personnel records) can be estimated on a 

case-by-case basis using the approach described in ORAUT-TKBS-0007-5. 

Based on the procedural information and the data on-hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to other 

radionuclides for CFA workers during the period from January 1, 1953 through December 31, 1968. 

7.2.7 Internal Dose Reconstruction Feasibility Conclusion

The 1967-1970 period at INL has been identified as an area requiring analysis for the entire site due to 

a change in bioassay protocol during that period. While a data-sufficiency issue has not been 

identified (with the exception of the period from 1963-1974 at CPP and the reserved sections/areas of 

this report), NIOSH has determined that a bioassay co-worker model is needed to support 

reconstructing internal dose for the post-1966 period at INL. 

NIOSH has come to the following conclusions regarding the feasibility of internal dose reconstruction 

for the six operating areas under evaluation for this report: 

• Test Reactor Area: NIOSH concludes that the available bioassay data are such that internal 

radiation doses received from intakes of fission products and actinides can be completely 

reconstructed with sufficient accuracy for the Test Reactor Area at INL from 1953 through 1970. 

• Chemical Processing Plant: NIOSH concludes that the available bioassay data are such that 

internal radiation doses received from intakes of fission products and other radionuclides can be 

completely reconstructed with sufficient accuracy for the Chemical Processing Plant at INL from 

1953 through 1962. Increased potential for intake due to poor contamination control and 

inadequate personnel monitoring for exposures to transuranics separated from mixed fission 

products makes it unlikely that exposures to alpha emitters can adequately be reconstructed from 

January 1963 through December 1974. 

NOTE: Based on a need to further evaluate the implementation of the CPP programmatic 

improvements, NIOSH has reserved the time period beginning January 1, 1975 for further 

evaluation. 
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• Test Area North: Based on the availability of the gross beta urine sample data, gross gamma urine 

sample data, and whole-body count data for TAN workers, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses for TAN workers 

during the period from the onset of TAN operations in 1955 through 1970. 

NOTE: The only exception to the above conclusion is potential uranium exposures that might 

have occurred without mixed fission products being present. Those potential exposures 

might have occurred during 1961–1970, and are being reserved for further evaluation. 

• Misc. Reactor Areas: NIOSH concludes that the available bioassay data are such that internal 

radiation doses received from intakes of mixed fission and activation products, and other 

applicable radionuclides, can be completely reconstructed with sufficient accuracy for the 

Miscellaneous Reactor Areas at INL from 1955 through 1970. 

NOTE: The year 1968 is being reserved to further investigate Pa-233 separations at the ARA-I hot 

cells.

• Central Facilities Area: NIOSH concludes that the available bioassay data are such that internal 

radiation doses received from intakes of mixed fission and activation products, and other 

applicable radionuclides, can be completely reconstructed with sufficient accuracy for the Central 

Facilities Area at INL from 1949 through 1970. 

• Burial Ground: NIOSH concludes that the available bioassay data are such that internal radiation 

doses received from intakes of mixed fission and activation products, and other applicable 

radionuclides, can be completely reconstructed with sufficient accuracy for the Burial Ground 

from 1953 through 1968. 

NOTE: The years 1969 forward are being reserved to further investigate drum-retrieval activities.

Although NIOSH found that it is not possible to completely reconstruct internal radiation doses for 

workers in the Chemical Processing Plant at the Idaho National Laboratory for the period from 

January 1, 1963 through December 31, 1974, NIOSH intends to use any internal monitoring data that 

may become available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Dose reconstructions for individuals working in the CPP

during the period from January 1, 1963 through December 31, 1974, but who do not qualify for 

inclusion in the SEC, may be performed using these data, as appropriate. 
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7.3 Evaluation of Bounding External Radiation Doses at INL

The principal source of external radiation doses for INL workers was a combination of direct 

exposures from nuclear reactors and exposures to the various radionuclide source terms at the site, 

which were predominately attributed to exposures to mixed fission products, activation products, and 

uranium. See Section 5.0 for a more detailed discussion of the radionuclides that are known to have 

been present. 

The following subsections address the ability to bound external doses, methods for bounding doses, 

and the feasibility of external dose reconstruction. 

7.3.1 Evaluation of Bounding Process-Related External Doses

The following subsections summarize the extent and limitations of information available for 

reconstructing the process-related external doses of members of the class under evaluation.

7.3.1.1 Electron and Photon Dosimetry Data

Because all workers in the INL’s radiological areas were monitored for exposures to electron and 

photon radiation, and because the DOE provides workers’ dosimeter data to NIOSH for dose 

reconstructions, the external electron and photon doses for the monitored workers can be 

reconstructed using the guidance in ORAUT-TKBS-0007-6. Because the control dosimeters at INL 

were likely exposed to elevated levels of onsite ambient (environmental) external radiation due to 

being located at the entrances of the major operating areas, onsite ambient external doses also need to 

be assigned to all monitored INL workers to account for any radiation dose that may have been 

inappropriately subtracted from their dosimeter results as background radiation (ORAUT-TKBS0007-6). 

For unmonitored workers, onsite ambient (environmental) external doses may be assigned. The onsite 

ambient external doses are limited to photon doses because the unmonitored workers would not have 

been close enough to the radioactive source terms to receive an external electron dose. 

7.3.1.2 Neutron Dosimetry Data

The majority of the INL workers were not monitored for neutron dose. However, the available 

information indicates that neutron doses were unlikely at INL and that neutron radiation fields are 

specific to a few facilities (ORAUT-TKBS-0007-6). During the period of evaluation, ORAUTTKBS-0007-6 indicates that the only location where neutron exposures were possible were at the 

MTR, TRA Hot Cell Cave, CFA calibration facilities (CF-633 and CF-636), and near some of the 

transuranics at the Burial Ground (ORAUT-TKBS-0007-6). Recently-captured information, such as 

the references for Section 5.0, indicates that neutron calibration/check sources (e.g., Ra-Be, Po-Be, 

Pu-Be, Am-Be, Cf-252) were present at most of the major operating areas at one time or another. 

Therefore, anyone working near those sources while they were unshielded would have also been 

exposed to neutron radiation. 
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Informal observations\reviews of the INL’s neutron dosimetry data, conducted while performing dose 

reconstructions and while capturing records at the INL site, indicate that the neutron doses for the 

monitored workers are typically less than the dosimeters’ limits of detection. In the INL claims, 

career neutron doses of 500 mrem or greater are rare. The available information also indicates that the 

INL investigated neutron exposures to unmonitored workers and estimated doses for those workers. 

As an example, neutron dose was estimated for an MTR worker after a neutron beam was discovered 

in the area where he was working (Sommers, 1957). Given that the INL likely monitored the workers 

with the highest potential to receive neutron doses, and given that the monitored workers’ neutron 

doses were typically less than the dosimeters’ limits of detection, it is unlikely that an unmonitored 

INL worker received more than an incidental exposure to neutron radiation. 

Because unmonitored neutron doses were unlikely at the INL, and because DOE provides the 

workers’ dosimeter data to NIOSH for dose reconstructions, neutron doses for the workers can be 

reconstructed using the guidance in ORAUT-TKBS-0007-6. 

7.3.2 Evaluation of Bounding Ambient Environmental External Doses

Table 4-13 of ORAUT-TKBS-0007-4 provides the onsite ambient (environmental) external doses 

from 1952 through the end of the evaluation period. For the years 1949–1951, the onsite ambient 

doses can be bounded by assigning the onsite ambient doses being assigned for the years 1952–1972 

because radiological operations prior to 1952 were minimal compared to 1952 and later (ORAUTTKBS-0007-2). 

Because all workers in the INL’s radiological areas were monitored for external dose, and because 

DOE provides workers’ dosimeter data to NIOSH for dose reconstructions, the only doses that need to 

be assigned for unmonitored workers at INL are onsite ambient external doses. Because the control 

dosimeters at INL were likely exposed to elevated levels of onsite ambient radiation due to being 

located at the entrances of the major operating areas, onsite ambient doses are also assigned to all 

monitored INL workers to account for any radiation dose that may have been inappropriately 

subtracted from their dosimeter results as background radiation (ORAUT-TKBS-0007-6). To bound 

any onsite ambient dose when an energy employee’s work location for a given year is unknown 

(which is often the case for unmonitored workers), the highest onsite ambient dose for any operating 

area can be assigned for that year. 

7.3.3 Evaluation of Bounding Occupational Medical Doses 

NIOSH has sufficient information through documents in the SRDB and ORAUT-TKBS-0007-3 to 

support the ability to bound the occupational medical doses for the evaluated class. 
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7.3.4 Methods for Bounding External Dose at INL

There is an established protocol for assessing external exposure when performing dose reconstructions 

(these protocol steps are discussed in the following subsections):

• Electron and Photon Doses 

• Neutron Dose 

• Ambient Environmental External Dose 

• Occupational Medical Dose

7.3.4.1 Methods for Bounding Operational Period External Dose

Electron and Photon Doses

NIOSH has determined that external electron and photon doses can be reconstructed. 

The primary method is the use of individual’s external dosimetry records that are provided by DOE. 

Dosimeter exchange frequencies and thresholds are known and documented. Missed dose may be 

applied using that information. The necessary biases/corrections that need to be made to the worker’s 

electron and photon dosimetry results are provided in ORAUT-TKBS-0007-6. To account for 

unmonitored radiation exposures, onsite ambient external doses can be assigned in accordance with 

the guidance in ORAUT-TKBS-0007-4 and ORAUT-PROC-0060. 

Neutron Dose

NIOSH has determined that external neutron doses can be reconstructed. 

The primary method is the use of individual’s external dosimetry records that are provided by DOE. 

Dosimeter exchange frequencies and thresholds are known and documented. Missed neutron dose 

may be applied using that information. The necessary biases/corrections that need to be made to the 

worker’s neutron dosimetry results are provided in ORAUT-TKBS-0007-6. Because unmonitored 

neutron doses were unlikely at INL, unmonitored neutron doses do not need to be assigned. 

Occupational Medical Dose 

NIOSH has determined that occupational medical doses can be reconstructed. 

NIOSH has sufficient information through documents in the SRDB and ORAUT-TKBS-0007-3 to 

support the ability to bound the occupational medical doses for the evaluated class. 

7.3.4.2 Methods for Bounding Ambient Environmental External Doses

NIOSH has determined that ambient environmental external doses can be reconstructed. The ambient 

environmental external doses can be reconstructed using the guidance in ORAUT-TKBS-0007-4 and 

ORAUT-PROC-0060. 
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7.3.5 External Dose Reconstruction Feasibility Conclusion 

NIOSH has reached the following conclusions regarding the feasibility of external dose 

reconstruction: 

• External dose for external electron and photon exposures may be reconstructed for all members of 

the evaluated INL class during the period under evaluation. 

• External dose for neutron exposures may be reconstructed for all applicable members of the 

evaluated INL class during the period under evaluation. 

• Occupational medical dose for occupational X-ray exposures may be reconstructed for all 

members of the evaluated INL class for the period under evaluation. 

7.4 Evaluation of Petition Basis for SEC-00219

The following assertion was made on behalf of petition SEC-00219 for the Idaho National 

Laboratory: 

SEC-00219: There was no internal monitoring for plutonium, neptunium, or fission products at INL 

between 1949 and 1970. 

NIOSH has determined that internal monitoring of personnel has focused on fission products since the 

beginning of radiological operations at the Idaho National Laboratory. In-vitro bioassay began in 

1953 with 455 gross beta and gross beta/gamma analyses and a single uranium analysis. In 1961, 

whole-body counting marked the beginning of in-vivo bioassay for gamma-emitting radionuclides that 

are common in fission products. In-vitro and in-vivo measurements by year are located in Section 6.1

of this evaluation report. During the course of this evaluation, it was determined that internal 

monitoring for plutonium and neptunium, while limited, was adequately performed. However, 

beginning in 1963, increased alpha contamination levels at the Chemical Processing Plant were 

routinely detected with no subsequent increase in plutonium bioassay. Due to inadequate bioassay, 

NIOSH is recommending the period from January 1, 1963 through December 31, 1974 at CPP for 

inclusion in the SEC. Significant programmatic improvements were identified and initiated in 1974. 

Based on a need to further evaluate the implementation of these improvements, NIOSH has reserved 

the time period beginning January 1, 1975 for further evaluation (ICPP HP Upgrade, 1971-1980). 

With the exception of this proposed class period, and the locations and periods reserved for further 

evaluation, NIOSH finds there are sufficient data and information relating to internal exposures and 

monitoring for the period between 1949 and 1970 at INL. 
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7.5 Other Potential SEC Issues Relevant to the Petition Identified During the 

Evaluation 

During the feasibility evaluation for SEC-00219, a number of issues were identified that needed 

further analysis and resolution. The issues and their current status are:

• ISSUE 1: SEC-00172 Petitioner-Proposed SEC Petition Class: All employees who worked in all 

facilities and areas that have been or are operated by the DOE at Idaho National Laboratory, 

Scoville, Idaho, from January 1, 1950 through December 31, 2005. 

RESPONSE: NIOSH completed the SEC-00172 PDR-Professional Judgment with the following 

finding: Based on the information provided by the petitioner and documentation available, NIOSH 

finds that there is no support for the petition bases. The petition does not demonstrate: 

1. There exists one or more unmonitored, unrecorded, or inadequately monitored or 

recorded exposure incidents;

2. Exposures and radiation doses potentially incurred by members of the proposed class 

were not monitored, either through personal monitoring or through area monitoring; 

3. Radiation monitoring records were lost, falsified, or destroyed; or that there is no 

information regarding monitoring, source, source term, or process from the site where 

the employees worked. 

As part of the assessment for this SEC-00219 INL SEC evaluation, NIOSH has reviewed the 

information and findings of the SEC-00172 petition review and PDR-Professional Judgment and 

incorporated the applicable information into this evaluation report. 

• ISSUE 2: Previous petitions evaluated for INL – from Completed ER Matrix

RESPONSE: SEC-00172 is the only previous petition for INL. As part of the assessment for this 

INL SEC evaluation, NIOSH has reviewed the information and findings of the SEC-00172 

petition review and PDR-Professional Judgment and incorporated the applicable information into 

this evaluation report.

• ISSUE 3: The SEC-00219 Petitioner requests that the evaluated class include all employees who 

worked in any location at the Idaho National Laboratory from January 1, 1949 to December 31, 

1970. 

RESPONSE: The SEC-00219 Evaluation Report evaluated all employees who worked at the 

Idaho National Laboratory from January 1, 1949 to December 31, 1970. Argonne National 

Laboratory-West, located on the INL reservation, will be evaluated as a separate site, per the 

requirements of the EEOICPA SEC evaluation process, under its own SEC petition number 

(SEC-00224). 
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• ISSUE 4: Recent concerns have been raised by the ABRWH about potential actinide exposures at 

the INL-Radioactive Waste Management Complex.

RESPONSE: The ABRWH concerns about potential actinide exposures at the Radioactive Waste 

Management Complex pertain to waste retrieval. In 1969, the first exhumation and retrieval 

project took place at the Burial Ground (also referred to as the Subsurface Disposal Area of the 

Radioactive Waste Management Complex). At the time of this SEC evaluation, sufficient 

information relating to the post-1968 waste retrieval in the Burial Ground is not available to 

support a feasibility conclusion in this report. Accordingly, the post-1968 period at the Burial 

Ground has been reserved for additional evaluation, which will be included as an Addendum to 

this evaluation report, or as a separate 83.14 SEC evaluation if the research leads to the conclusion 

that the current proposed INL SEC class should be extended. 

• ISSUE 5: CPP: Assess the ability to reconstruct dose for the portion of the class from 1963-1974 

(recommending a SEC class for CPP from 1/63-12/74) 

RESPONSE: NIOSH has assessed the CPP exposure sources and associated dose reconstruction 

methods in this report. Increased potential for intake due to poor contamination control and 

inadequate personnel monitoring for exposures to transuranics separated from mixed fission 

products makes it unlikely that exposures to alpha emitters can adequately be reconstructed from 

January 1963 through December 1974. An SEC class is recommended for CPP from January 1, 

1963 to December 31, 1974 due to degradation in radiological conditions resulting in increased 

alpha contamination without adequate bioassay. 

• ISSUE 6: CPP: Assess the ability to reconstruct dose for the portion of the class beyond 1974. 

RESPONSE: At the time of this SEC evaluation, the end date for proposed SEC class at CPP has 

been defined as 1974. Accordingly, the post-1974 period at CPP has been reserved for additional 

evaluation, which will be included as an Addendum to this evaluation report or as a separate 83.14 

SEC evaluation if the research leads to the conclusion that the current proposed INL SEC class 

should be extended. 

• ISSUE 7: Need for a mixed-fission-products co-worker model beginning in 1967 due to a change 

in the in-vivo counting protocol. 

RESPONSE: In 1967 INL transitioned towards more reliance on in-vivo radiobioassay resulting in 

fewer in-vitro radiobioassay analyses. At that time INL also began a representative in-vivo

counting program such that workers were only counted every four years. The ability to bound 

doses requires an in-vitro coworker model.

• ISSUE 8: V&V of bioassay data.

RESPONSE: It is firmly believed that the in-vivo and in-vitro bioassay datasets for INL are 

complete. However, a verification and validation has not been finished at this time and will be 

performed as part of the development of a future internal co-worker study for INL. 
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• ISSUE 9: Use dosimeter information to place workers in particular areas.

RESPONSE: NIOSH has determined that because all of the workers in INL’s radiological areas 

were monitored for external exposures, and the records exist for those workers, this information 

coupled with the internal monitoring records can be used to place workers in particular 

radiological work areas at INL. This further supports the ability to separate the CPP as an 

individual class at INL for the purposes of this SEC evaluation. In addition, a bioassay co-worker 

model will be developed due to a change in in-vivo and in-vitro bioassay practices in 1967. 

NIOSH believes that a bioassay co-worker model stratified by area can be developed based on 

personnel location information from dosimeter records. 

• ISSUE 10: Reserve TAN-615 Actuator Building 1961-1970 

RESPONSE: At the time of this SEC evaluation, sufficient information relating to TAN-615, 

Actuator Building, from 1961-1970 is not available to support a feasibility conclusion in this 

report. Accordingly, the 1961-1970 period at TAN-615 has been reserved for additional 

evaluation, which will be included as an Addendum to this evaluation report or as a separate 83.14 

SEC evaluation if the research leads to the conclusion that the current proposed INL SEC class 

should be extended. 

• ISSUE 11: Reserve ARA-I in 1968 due to Pa-233 separation. 

RESPONSE: At the time of this SEC evaluation, sufficient information relating to Pa-233 

exposures at ARA-1 in 1968 is not available to support a feasibility conclusion in this report. 

Accordingly, the 1968 period at ARA-1 has been reserved for additional evaluation, which will be 

included as an Addendum to this evaluation report, or as a separate 83.14 SEC evaluation if the 

research leads to the conclusion that the current proposed INL SEC class should be extended. 

• ISSUE 12: OMRE

RESPONSE: During research for the SEC-00219 evaluation report, it was discovered that the 

Organic Moderated Reactor Experiment (OMRE) at INL was under the direction of the AEC 

Chicago Operations Office and not the AEC Idaho Operations Office. OMRE was not part of 

Argonne National Laboratory-West, nor is it a separate covered facility. NIOSH does not believe 

that this impacts the evaluation decisions presented in this report; a subsequent review of this 

analysis will be presented separately. 

• ISSUE 13: In the evaluation of the 1963 TRA plutonium contamination incident, documentation 

was found that indicated a link between the plutonium material and CPP (Sommers, 1963). A 

chemist from CPP filled 13 capsules with rare-earth oxides, fission monitors, and transuranic

oxides by weight in the CPP X-Cell Laboratory. Because the samples were isotopically pure, an 

investigation was performed by NIOSH to determine whether isotopic separations were performed 

at CPP. 
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RESPONSE: The materials involved in the 1963 TRA contamination incident were measured into 

capsules within the CPP X-Cell, but were not separated at this facility. Documentation indicates 

that the materials were ordered from ORNL and delivered with their specified isotopic purities 

(Shank, 1964). With this knowledge, NIOSH considers this issue resolved. 

• ISSUE 14: The Umpire Laboratory Program began in 1968 with the aim of qualifying U.S. and 

foreign laboratories to analyze uranium and plutonium materials for the AEC. The program used 

samples prepared from very pure, well-characterized source materials. As announced by the AEC 

on May 5, 1965, in Press Release IN-584, the Atomic Energy Division of Phillips Petroleum 

Company was selected to conduct the program (Tingey, 1968). Because the program involved the 

preparation and analysis of plutonium samples, NIOSH began an investigation into what role INL 

had in preparing the plutonium standards. 

RESPONSE: Standardizations of the plutonium nitrate, plutonium oxides, and plutonium metal 

standards were completed by Atlantic Richfield Hanford Company, Los Alamos, New Brunswick 

Lab, and Rocky Flats Plant. Test samples of these materials were distributed to five U. S. and five 

foreign laboratories. INL neither prepared nor analyzed the plutonium samples (Shank, 1969). 

NIOSH considers this issue closed. 
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7.6 Summary of Feasibility Findings for Petition SEC-00219

This report evaluates the feasibility for completing dose reconstructions for employees at the Idaho 

National Laboratory from January 1, 1949 through December 31, 1970. NIOSH found that the 

available monitoring records, process descriptions and source term data available are not sufficient to 

complete dose reconstructions for the evaluated class of employees.

Table 7-17 summarizes the results of the feasibility findings at Idaho National Laboratory for each 

exposure source during the time period January 1, 1963 through December 31, 1974 (Chemical 

Processing Plant) and from January 1, 1949 through December 31, 1970 (remainder of INL site, 

including CPP from 1953-1962, and excluding the reserved portions of the site). 

Table 7-17: Summary of Feasibility Findings for SEC-00219

January 1, 1963 through December 31, 1974 (Chemical Processing Plant); 

January 1, 1949 through December 31, 1970 (INL site excluding CPP SEC class and reserved areas)

Source of Exposure

Chemical Processing Plant

Jan. 1, 1963 through Dec. 31, 1974

INL Site Excluding CPP SEC

Class and Reserved Areas

Jan. 1, 1949 through Dec. 31, 1970

Reconstruction 

Feasible

Reconstruction 

Not Feasible

Reconstruction 

Feasible

Reconstruction 

Not Feasible

Internal1 X2 X 

 - U X X

 - Np X X

 - Pu X X

 - Th X X

 - Other transuranic 

radionuclides

X X

 - Fission products X X

 - Noble gases X X

 - Iodine X X

 - Tritium X X

External X3 X 

 - Gamma X X

 - Beta X X

 - Neutron X X

- Occupational Medical X-ray X X 1 Internal evaluation of bioassay data included urinalysis/fecal (in-vitro) data and lung/whole body count (in-vivo) data

PARTIAL DOSE RECONSTRUCTION INFORMATION: 

2 INTERNAL: For the purpose of partial dose reconstruction for members of the proposed SEC class who worked at 

CPP from 1963 through 1974, partial DRs could include dose from any radionuclides for which specific personal 

monitoring data exist that can be assessed by methods defined in approved NIOSH dose reconstruction procedures or 

processes, but may not include dose from U, Np, Pu, or Th (or other alpha-emitters) that have been separated from 

fission products, based on ratios of these with fission products.
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3 EXTERNAL: For the purpose of partial dose reconstruction for members of the proposed SEC class who worked at 

CPP from 1963 through 1974, partial DRs should include external dose based on the specific personal monitoring data 

that exist, or external DR methods as defined in approved NIOSH dose reconstruction procedures or processes.

As of February 2, 2015, a total of 1074 claims have been submitted to NIOSH for individuals who 

worked at INL during the period under evaluation in this report. Dose reconstructions have been 

completed for 956 individuals (~98%). 

Although NIOSH found that it is not possible to completely reconstruct radiation doses for the 

proposed class, NIOSH intends to use any internal and external monitoring data that may become 

available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Therefore, dose reconstructions for individuals employed in 

the Chemical Processing Plant at the Idaho National Laboratory during the period from January 1, 

1963 through December 31, 1974, but who do not qualify for inclusion in the SEC, may be performed 

using these data as appropriate.

Table 7-18 provides a more general summary of the six individual feasibility tables provided in 

Section 7.2, Evaluation of Bounding Internal Radiation Doses at INL (i.e., Sections 7.2.1 through 

7.2.6 for each of the six operating areas under evaluation). Whereas the individual feasibility tables 

are more specific as to exposure sources, Table 7-18 provides the most restrictive feasibility 

designation for each year for that operating area.

The reserved portions of the site are: 

• TAN: TAN-607 (Fuel Storage Vaults) and TAN-615 (Actuator Building) from 1961-1970 due to 

potential uranium exposures without mixed fission products being present 

• Misc. Reactor Areas: ARA-I in 1968 due to potential unmonitored exposure to Pa-233

• Burial Ground: The years 1969 forward due to a newly-implemented procedure of waste 

exhumation and retrieval 
SEC-00219 03-12-15 Idaho National Laboratory

Table 7-18: Summary of INL Feasibility Determinations by Operating Area (1949-1974)

INL Operating Area

Years

49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74

Test Reactor Area N N N F F F F F F F F F F F F F F F C C C C N N N N 

Chemical Processing Plant N N N N F F F F F F F F F F I I I I I I I I I I I I 

Test Area North N N N N N N F F F F F F R R R R R R R R R R N N N N 

Misc. Reactor Areas N N N N N N F F F F F F F F F F F F C R C C N N N N 

Central Facilities Area F F F F F F F F F F F F F F F F F F C C C C N N N N 

Burial Ground N N N F F F F F F F F F F F F F F F C C R R N N N N 

F = Dose reconstructions are feasible.

C = Dose reconstructions are feasible but a mixed-fission-product co-worker model is needed.

I = Dose reconstructions are infeasible.

R = Reserved for further evaluation.

N = Prior to radiological operations or outside of evaluation period

220 of 258 
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8.0 Evaluation of Health Endangerment for Petition SEC-00219

The health endangerment determination for the class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(3). Under these requirements, if it is not 

feasible to estimate with sufficient accuracy radiation doses for members of the class, NIOSH must 

also determine that there is a reasonable likelihood that such radiation doses may have endangered the 

health of members of the class. Section 83.13 requires NIOSH to assume that any duration of 

unprotected exposure may have endangered the health of members of a class when it has been 

established that the class may have been exposed to radiation during a discrete incident likely to have 

involved levels of exposure similarly high to those occurring during nuclear criticality incidents. If 

the occurrence of such an exceptionally high-level exposure has not been established, then NIOSH is 

required to specify that health was endangered for those workers who were employed for a number of 

work days aggregating at least 250 work days within the parameters established for the class or in 

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

employees in the SEC. 

NIOSH has determined that the increased potential for intake due to poor contamination control and 

inadequate personnel monitoring for exposures to transuranics separated from mixed fission products 

makes it unlikely that exposures to alpha emitters can adequately be reconstructed from January 1963 

through December 1974. The Preliminary ICPP Health Physics Upgrade Program report was 

published and delivered to CPP management in October 1974. The class end date was changed to 

December 31, 1974 as a convenience since further NIOSH-ORAUT review and assessment will be 

necessary to determine if the program data from 1975 forward is adequate to support reconstructing 

dose with sufficient accuracy for this area.

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

establish that members of the proposed SEC class at the Idaho Chemical Processing Plant (CPP) 

between January 1, 1963 and December 31, 1974 were exposed to radiation during a discrete incident 

resulting in significant unmonitored exposures likely to have involved exceptionally high-level 

exposures. However, evidence indicates that some workers in the proposed class may have 

accumulated substantial chronic exposures through episodic intakes of radionuclides, combined with 

external exposures to gamma, beta, and neutron radiation. Based on its assessment, presented in this 

evaluation report, NIOSH finds that there were issues that make it unlikely that exposures to 

alpha-emitters can be adequately reconstructed from January 1, 1963 through December 31, 1974. 

Consequently, NIOSH has determined that health was endangered for those CPP workers from 

January 1, 1963 through December 31, 1974 who were employed for at least 250 aggregated work 

days either solely under their employment or in combination with work days within the parameters 

established for other SEC classes.

For the workers at the INL site for the period of January 1, 1949 through December 31, 1970 

(excluding the CPP workers from January 1, 1963 through December 31, 1974, and those workers in 

the areas with reserved periods in this report), a health endangerment determination is not required 

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

the evaluated class.
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9.0 Class Conclusion for Petition SEC-00219

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

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class to be added to the SEC includes all employees of the Department of Energy, its 

predecessor agencies, and their contractors and subcontractors who worked at the Idaho National 

Laboratory in Scoville, Idaho, and were monitored for external radiation at the Idaho Chemical 

Processing Plant (CPP) (e.g., at least one film badge or TLD dosimeter from CPP) between January 1, 

1963 and December 31, 1974 for a number of work days aggregating at least 250 work days, 

occurring either solely under this employment, or in combination with work days within the 

parameters established for one or more other classes of employees in the Special Exposure Cohort. 

For all non-CPP workers on the remainder of the INL site, there are sufficient monitoring data to 

make dose reconstruction feasible for the time period from January 1, 1949 through December 31, 

1970. 

NIOSH has carefully reviewed all material sent in by the petitioner, including the specific assertions 

stated in the petition, and has responded herein (see Section 7.4). NIOSH has also reviewed available 

technical resources and many other references, including the Site Research Database (SRDB), for 

information relevant to SEC-00219. In addition, NIOSH reviewed its NOCTS dose reconstruction 

database to identify EEOICPA-related dose reconstructions that might provide information relevant to 

the petition evaluation.

These actions are based on existing, approved NIOSH processes used in dose reconstruction for 

claims under EEOICPA. NIOSH’s guiding principle in conducting these dose reconstructions is to 

ensure that the assumptions used are fair, consistent, and well-grounded in the best available science. 

Simultaneously, uncertainties in the science and data must be handled to the advantage, rather than to 

the detriment, of the petitioners. When adequate personal dose monitoring information is not 

available, or is very limited, NIOSH may use the highest reasonably possible radiation dose, based on 

reliable science, documented experience, and relevant data to determine the feasibility of 

reconstructing the dose of an SEC petition class. NIOSH contends that it has complied with these 

standards of performance in determining the feasibility or infeasibility of reconstructing dose for the 

class under evaluation. 
SEC-00219 03-12-15 Idaho National Laboratory

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10.0 References

42 C.F.R. pt. 81, Guidelines for Determining the Probability of Causation Under the Energy 

Employees Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol. 67, No. 85/Thursday, p. 22,296; May 2, 2002; SRDB Ref ID: 19391 

42 C.F.R. pt. 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule; May 2, 2002; SRDB Ref ID: 

19392 

42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as Members of the Special 

Exposure Cohort Under the Energy Employees Occupational Illness Compensation Program Act of 

2000; Final Rule; May 28, 2004; SRDB Ref ID: 22001

42 U.S.C. §§ 7384-7385 [EEOICPA], Energy Employees Occupational Illness Compensation 

Program Act of 2000, as amended

AEC, 1954-1960, Various INEL Area Maps; U.S. Atomic Energy Commission (AEC) Idaho 

Operations Office; various dates ranging from 1954 to 1960; SRDB Ref ID: 117427 

AEC, 1954-1986, National Reactor Testing Station Plot Plans for Various Areas; U.S. Atomic Energy 

Commission (AEC) Idaho Operations Office; various dates ranging from 1954 to 1986; SRDB Ref 

ID: 86891 

AEC, Jan-Dec1967, Major Activities in the Atomic Energy Programs, January-December 1967, 

INDC-216; U.S. Atomic Energy Commission; January 1968; SRDB Ref ID: 115844 

AEC, Apr1967, Quarterly Technical Report STEP Project for July 1966 – September 1966, IDO17213; U.S. Atomic Energy Commission (AEC); April 1967; SRDB Ref ID: 140006 

AEC, 1970, Nuclear Technology Branch Annual Progress Report for Period Ending June 30, 1969, 

IN-1317; Atomic Energy Commission (AEC); Issued January 1970; SRDB Ref ID: 136165 

AEC Annual Report, 1960, Annual Report to Congress of the Atomic Energy Commission for 1960, 

U.S. Atomic Energy Commission; January 1961; SRDB Ref ID: 40909 

Affidavit, 2014, Affidavit from Petitioner supporting Petition SEC-00219 for Idaho National 

Laboratory, received August 29, 2014; DSA Ref ID: 120226 

Air Monitoring, Feb1953-Oct1954, Alpha Activity Results for Constant Air Monitors (February 1953-

October 1954); dates ranging from February 1953-October 1954; SRDB Ref ID: 138649 

Air Monitoring, Dec1956-Aug1957, Long Life Activity of Constant Air Monitors and 1⁄4 of Filters and 

Beta and Alpha Activity Results for Constant Air Monitors (December 1956-August 1957); dates 

ranging from December 1956-August 1957; SRDB Ref ID: 138651 
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Air Monitoring, 1965-1966, Air Samples, Calculations, and Plots (1965-1966); dates ranging from 

1965-1966; SRDB Ref ID: 138630 

Air Monitoring, Mar1965-Dec1966, MTR Air Monitoring Surveys and Isotopic Analysis (March 

1965-December 1966); dates ranging from March 1965-December 1966; SRDB Ref ID: 140302 

Air Monitoring, 1966, MTR Stack, Particulate Spectrums, Isotopic Analysis and Air Monitoring 

(1966); various dates throughout 1966; SRDB Ref ID: 140286 

Air Monitoring, 1966-1967, ETR High Volume Air Sample Spectra (1966-1967); dates ranging from 

1966-1967; SRDB Ref ID: 140177 

Alpha Surveys, 1962-1968, Test Area North Alpha Surveys; author unknown; 1962-1968; SRDB Ref 

ID: 140265 

Analytical Lab, 1957, Analytical Control Laboratory Operating Manual, IDO-14396; M. J. Painter; 

Phillps Petroleum Co.; March 21, 1957; SRDB Ref ID: 139890 

Analytical Program, 1957, Analytical Program for Processing Aluminum-Uranium Reactor Fuel 

Elements, IDO-14412, R. C. Shank; Phillips Petroleum Co.; August 15, 1957; SRDB Ref ID: 140043 

Anderson, 1974, Personnel Dosimetry and Exposure Control at the ICPP; D. K. Jenson; 1974; SRDB 

Ref ID: 118886, PDF p. 8

Becar, 1961, Critical Experiment Tank (CET) Reactor Hazards Summary, XDC-61-4-709; N. J. 

Becar, J. F. Kunze, G. D. Pincock, and L. D. VanVleck; March 31, 1961; SRDB Ref ID: 1093 

Belnap, 1963, Waste Disposal Report and Authorization Form for CFA Laundry; October 1963; Vern 

E. Belnap; SRDB Ref ID: 139214 

Bengstad, 1962, Radioactive Shipment Record; R. H. Bengstad; October 16, 1962; SRDB Ref ID: 

138788, PDF p. 17 

Bentzen, 1963, Safety Analysis Report – SNAPTRAN 2/10A-1 Safety Tests, IDO-16825; F. L. Bentzen 

et al.; January 28, 1963; SRDB Ref ID: 136849 

Bioassay, 1956-1984, INEL Bioassay CD 1 of 3; author unknown; 1956-1984; SRDB Ref ID: 123095 

Bioassay, 1956-1967, INEL Bioassay CD 2 of 3; author unknown; 1956-1957; SRDB Ref ID: 123096 

Bioassay, 1958-1986, INEL Bioassay CD 3 of 3; author unknown; 1958-1986; SRDB Ref ID: 123104 

Bjorklund, 1970, Revised Safety Analysis Report for the ICPP Denitration Process; W. J. Bjorklund 

and G. F. Offutt; December 11, 1970; SRDB Ref ID: 116061 
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Boone, 1959, Gross Activity Stack Monitoring Results and Environmental Hazards Evaluation for the 

D101-L2A and D101-L2C Test Series, DC-59-7-722; F. W. Boone; July 30, 1959; SRDB Ref ID: 

31951 

Brugger, 1975, Nuclear Technology Division Annual Progress Report for Period Ending June 30, 

1974, ANCR-1177; R. M. Brugger; January 1975; SRDB Ref ID: 140199, PDF p. 24 

Burial Ground, 1962, NRTS Burial Ground Standard Practices Manual; 5.61.2; author unknown; 

SRDB Ref ID 140874, PDF p. 3 

CAM, 1961, CAM Filter Counts for Long-Lived Alpha and Beta 1961; author unknown; 1961; SRDB 

Ref ID: 140193 

CAM, 1962-1963, Test Area North Logbook of CAM Filter Results 1962-1963; 

Author unknown; 1962-1963; SRDB Ref ID: 140255 

CAM, 1963-1964, Test Area North Logbook of CAM Filter Results 1963-1964; author unknown; 

1963-1964; SRDB Ref ID: 140257 

CAM, 1964-1967; Test Area North Logbook of CAM Filter Results 1964-1967; author unknown; 

1964-1967; SRDB Ref ID: 140259 

CAM 1967-1970, CAM Filter Counts from May 8, 1967 to March 12, 1970; unknown author; various 

dates; SRDB Ref ID: 138064 

CFA HP Logs, 1965-1966, CFA Health Physics Log Sheets January 1965 – December 1966; SRDB 

Ref ID: 139293 

Chemical Programs, 1972, Idaho Chemical Programs Annual Technical Report: Fiscal Year 1971, 

ICP-1006, J. A. Buckham; Allied Chemical Corporation; May 1972; SRDB Ref ID: 83365 

Cipperley, 1958, Personnel Metering Operating Techniques, Self-Service System, Effective August 20, 

1951 through March 29, 1958, F. V. Cipperley; IDO-12007; U.S. Atomic Energy Commission, Idaho 

Operations Office; 1958; SRDB Ref ID: 8112 

Cipperley, 1968, Personnel Dosimetry Standard Operating Procedure Effective March 30, 1958, F. 

V. Cipperley; IDO-12070; U.S. Atomic Energy Commission, Idaho Operations Office; December 

1968; SRDB Ref ID: 8086 

Condotta, 1981, Action Plans – Shift Laboratory; DC-3-81; D. L. Condotta; July 14, 1981; SRDB Ref 

ID: 124536, PDF p. 10 

Condotta, 1981a, Action Plans – Shift Laboratory; DC-3-81; D. L. Condotta; July 14, 1981; SRDB 

Ref ID: 124536, PDF p. 15

Conner, 1956, MTR Technical Branch Quarterly Report, First Quarter – 1956; W. P. Conner; June 8, 

1956; SRDB Ref ID: 139938 
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Cordes, 1963, TAN Health Physics Progress Reports for 1963, O. L. Cordes; various dates throughout 

1963; SRDB Ref ID: 138714 

Cordes, 1963a, TAN Health Physics Progress Report for March 1963, O. L. Cordes; April 23, 1963; 

SRDB Ref ID: 138714, PDF pp. 45-49 

CPP Air, 1957-1960, CPP Sample Record Sheets Air Samples February 1957 - January 1960; author 

unknown; SRDB Ref ID: 140233 

CPP Contract, 1955, Idaho Chemical Processing Plant Contract AT(10-1)-728, Design & Operations 

Manual for SIR-STR Fuel Processing BKC-4193; IDO-18002; January 3, 1955: SRDB Ref ID: 83201 

ECI

CPP Dosimetry, 1984, ICPP Internal Dosimetry Procedures Manual, L. Guinn; Westinghouse Idaho 

Nuclear Company, Inc.; 1984; SRDB Ref ID: 118948 

CPP Fact Sheet, undated, The Idaho Chemical Processing Plant (ICPP) (A Fact Sheet), B. F. 

Boardman; U. S. Atomic Energy Commission; undated; SRDB Ref ID: 14658 

CPP Hi Vol Air, 1960-1961, CPP High Volume Air Sample Logsheets with Alpha and Beta 

Measurements 1960-1961; author unknown; SRDB Ref ID: 140399 

CPP HP Manual, 1952, CPP Health Physics Manual April 1952; IDO-14055; author unknown; SRDB 

Ref ID: 8095 

CPP HP Surveys, 1970, CPP HP Routine Surveys December 1970; author unknown; 1970; SRDB Ref 

ID: 138084 

CPP Krypton, 1979, Recovery and Purification of Krypton Gas at ICPP, BUC-153-70; April 29, 

1979; SRDB Ref ID: 47090 

CPP Phase-out, 1994, Idaho Chemical Processing Plant and Plutonium-Uranium Extraction Plant 

Phase-out/Deactivation Study, WHC-EP-0693, M. W. Patterson; Westinghouse Hanford Company; 

January 1994; SRDB Ref ID: 88360 

CPP Plan, 1997, Field Sampling Plan for the D&D of the CPP-631 RaLa Building, and CPP-709 and 

CPP-734 Monitoring Stations at the Idaho Chemical Processing Plant, R. W. Jones; INEEL/EXT-97-

00805; Lockheed Martin; August 1997; SRDB Ref ID: 140747 

CPP Progress Report, Feb1954, Idaho Chemical Processing Plant Progress Report – February, 1955, 

F. P. Vance; IDO-14297; April 23, 1954; SRDB Ref ID: 87334 ECI

CPP Progress Report, Sep1955, Idaho Chemical Processing Plant Progress Report – September, 

1955, F. P. Vance; IDO-14355; October 24, 1955; SRDB Ref ID: 87332 ECI

CPP Progress Report, Nov1957, Idaho Chemical Processing Plant Progress Report – November, 

1957, F. P. Vance; PTR-262; December 12, 1957; SRDB Ref ID: 87154 ECI
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CPP Progress Report, Jan1958, Idaho Chemical Processing Plant Progress Report – January 1958, F. 

P. Vance; PTR-279; February 13, 1958; SRDB Ref ID: 83171 ECI

CPP Progress Report, Feb1958, Idaho Chemical Processing Plant Progress Report – February 1958, 

F. P. Vance; PTR-288; March 14, 1958; SRDB Ref ID: 88004 ECI

CPP Progress Report, Jul1958, Idaho Chemical Processing Plant Progress Report – July 1958, F. P. 

Vance; PTR-333; August 25, 1958; SRDB Ref ID: 83177 ECI

CPP Progress Report, Aug1958, Idaho Chemical Processing Plant Progress Report – August 1958, F. 

P. Vance; PTR-333; October 10, 1958; SRDB Ref ID: 83178 ECI

CPP Progress Report, Sep1958, Idaho Chemical Processing Plant Progress Report – September 1958, 

F. P. Vance; PTR-355; November 10, 1958; SRDB Ref ID: 83179 ECI

CPP Progress Report, Jun1959, Idaho Chemical Processing Plant Progress Report – June 1959, F. P. 

Vance; PTR-427; August 27, 1959; SRDB Ref ID: 83196 ECI

CPP RaLa, 1969, Control of Fission Product Activity During Short-Cooled Fuel Processing 

Connected with the ICPP RaLa Process, J. R. Bower; Idaho National Laboratory; paper presented in 

Vienna, Austria on May 27-30, 1969; SRDB Ref ID: 32034 

CPP Safety Review, 1974, Idaho Chemical Processing Plant Safety Review Document, Volume I, 

ICPP-SRD-1; G. K. Cederberg; Allied Chemical Corporation; March 1974; SRDB Ref ID: 14660 

CPP Waste Management, 1960, Waste Management at the Idaho Chemical Processing Plant, IDO14551, A. L. Ayers; Phillips Petroleum Company; December 5, 1960; SRDB Ref ID: 139906 

Cusimano, 1963, Latent Image Fading in Personal Neutron Monitoring Film at the NRTS, J. P. 

Cusimano; IDO-12031; U.S. Atomic Energy Commission, Idaho Operations Office; 1963; SRDB Ref 

ID: 8135 

D101 L2A-1, 1959, Test Results from D101 L2A-1 Power Testing, IET #14, DC 59-8-728; various 

authors; August 19, 1959; SRDB Ref ID: 12732 

Davis, 1970, Critical Experiments in an MTR Mockup Using Phoenix Fuel, BNWL-1481; E. C. 

Davis, Jr., R. I. Smith, L. D. Williams; June 1970: SRDB Ref ID: 118572 

DCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 1, 

National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; April 15, 2011; 

SRDB Ref ID: 94768

Dierks, 1963, Contamination in TAN-615 Building, correspondence to V. V. Hendrix; D. H. Dierks; 

July 3, 1963; SRDB Ref ID: 141264 

DOE, 1991, Idaho National Engineering Laboratory Historical Dose Evaluation, Volume 2: 

Appendices, DOE/ID-12119; U.S. Department of Energy (DOE) Field Office, Idaho; August 1991; 

SRDB Ref ID: 1109
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DOE, 2009, Engineering Evaluation/Cost Analysis for Decommissioning of the TRA-632 Hot Cells, 

DOE/ID-11397; Department of Energy (DOE); September 2009; SRDB Ref ID: 87764 

EG&G, 1977, LOFT Integral Test System Final Safety Analysis Report Revision 1, Volume 2 of 3; 

EG&G Idaho, Inc.; August 1977; SRDB Ref ID: 88511 

EOCR, 1959, Experimental Organic Cooled Reactor: Conceptual Design, J. R. Huffman; Phillips 

Petroleum Company; December 1, 1959; SRDB Ref ID: 137817 

ETR Exposure History, 1993, Occupational Radiation Exposure History of Idaho Field Office 

Operations at the INEL, J. R. Horan and J. B. Braun; EGG-CS-11143; Idaho National Engineering 

Laboratory; October 1993: SRDB Ref ID: 1023 

ETR Surveys, 1961-1962, ETR Cubicle Radiation Surveys 1961-1962; author unknown; 1961-1962; 

SRDB Ref ID: 140463

Fire, 1970, Burial Ground Fire June 1, 1970; various authors; 1970, SRDB Ref ID: 139276 

Freeman, 1964, Tan Hot Shop and Satellite Facilities, IDO-17032; R. A. Freeman; November 1964; 

SRDB Ref ID: 138791

G&H Air, 1955, G & H Cell (RAF) Air Samples; author unknown; 1955; SRDB Ref ID: 140227 

GCRE Final Report, 1964, Army Gas-Cooled Reactor Systems Program Final Report: Full Power 

and Limited Endurance Test of the ML-1 Nuclear Power Plant, IDO-28634, R. H. Chesworth; 

Aerojet-General Nucleonics; September 1964; SRDB Ref ID: 136848 

GCRE Progress Report, 1959, Army Gas-Cooled Reactor Systems Program, Monthly Progress 

Report, April 1959, G. A. Linenberger; Aerojet-General Nucleonics; May 25, 1959; SRDB Ref ID: 

136834 

General Electric, 1953, Nuclear Power Plant Testing in the IET, APEX-131; General Electric; May 

1953; SRDB Ref ID: 40238 

General Electric, 1961, Test Program and Procedures for STP Program A, single page; General 

Electric; July 1961; SRDB Ref ID: 1068 

Gesell, 1986, Minimum Reporting Levels for the Panasonic Dosimetry Systems, letter to dosimetry 

customers, T. F. Gesell; Radiological and Environmental Sciences Laboratory; May 15, 1986; SRDB 

Ref ID: 8123 

Gesell, 1992, Technical Basis of the INEL Personal Photon-Beta Dosimeter, T. F. Gesell, M. A. Hall, 

and B. L. Anderson; Idaho State University, Department of Physics, December 1, 1992; SRDB Ref 

ID: 8163 

Gesell, 1996, Technical Basis of the INEL Personal Neutron Dosimeter; Report 960112 Attachment 

#4; T. F. Gesell; January 12, 1996; SRDB Ref ID 8157, PDF p. 19 
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Ginkel, 1959, Report of Investigation, Nuclear Incident at the Idaho Chemical Processing Plant; 

William L. Ginkel; undated; SRDB Ref ID: 88098 

Ginkel, 1960, Nuclear Incident at the Idaho Chemical Processing Plant on October 16, 1959, Report 

of the Investigating Committee; William L. Ginkel; February 15, 1960; SRDB Ref ID: 12784 

Grigg, 2008, Safety Analysis Report for the INTEC Laboratory Facilities (CPP-602 Laboratory and 

CPP-630 Mass Spectrometry Facility), SAR-121; Document owner: Clayne B. Grigg; May 20, 2008; 

SRDB Ref ID: 116935

H&S Bulletin, 1958, H&S Instruction Bulletin 3.20: Neutron Sources; H&S IB-3.20; author 

unknown; June 10, 1958; SRDB Ref ID: 85432, PDF pp. 137-138 

Hawley, 1965, Alpha Contamination Incident at MTR; C. A. Hawley; September 14, 1965; SRDB Ref 

ID: 140170 

Hayden, 1959, CPP Incident – October 16, 1959; Ha-180-59A; R. E. Hayden; November 3, 1959; 

SRDB Ref ID: 31947 

Hayden, 1959a, CPP Health Physics Monthly Report, Period July 1 through July 31, 1959; Ha-135-

59A; R. E. Hayden; August 7, 1959; SRDB Ref ID: 87080, PDF p. 21 

Hayden, 1959b, CPP Health Physics Monthly Report, Period August 1 through August 31, 1959; Ha159-59A; R. E. Hayden, September 9, 1959; SRDB Ref ID: 87080, PDF p. 14 

Hayden, 1961, HP Accomplishments 1960; Ha-11-61A-N; R. E. Hayden; February 3, 1961; SRDB 

Ref ID: 139251 

Hayden, 1962, NRTS Burial Ground Standard Practices Manual; 5.61.2; R. E. Hayden; April 13, 

1962; SRDB Ref ID: 140874 

Hazards Control Qtr, 1965, Hazards Control Quarterly Report No. 22 (July-September 1965); UCRL14534; various authors; undated; SRDB Ref ID: 13888 

Hofmann, 1965, Progress Report – Phoenix Fuel Program – FY-1965, BNWL-152; P. L. Hofmann; 

August 15, 1965; SRDB Ref ID: 118622 

Hoover, 1961, SL-1 Health Physics Data Report No. II, DR-RO-2; R. L. Hoover; February 6, 1961; 

SRDB Ref ID: 83650

Horan, 1959, Health and Safety Division Annual Report, 1958, IDO-12012; John R. Horan and 

multiple other authors; not dated, but likely written in 1959; SRDB Ref ID: 8079

Horan, 1961, Health and Safety Division Annual Report, 1960, IDO-12019; John R. Horan and 

multiple other authors; September 1961; SRDB Ref ID: 8078 
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Horan, 1962, Health and Safety Division Annual Report, 1961, IDO-12021; John R. Horan and 

multiple other authors; May 21, 1962; SRDB Ref ID: 8076 

Horan, 1963, Health and Safety Division Annual Progress Report, 1963; IDO-12037; John R. Horan; 

August 1964; SRDB Ref ID: 13449, PDF p. 17 

Howe, 1986, Summary of B-331 Radiation Protection Program Results - 1985; H. J. Howe; March 3, 

1986; SRDB Ref ID: 14076 

HP Logbook, 1959, HP Logbook for SPERT August 12, 1958 – October 29, 1959; author unknown; 

1958-1959; SRDB Ref ID: 141077 

HP Log Sheets, 1972, Health Physics Log Sheets for June 1972-November 1972; SRDB Ref ID: 

141076 

HP Monthly Reports, 1957, Health Physics Monthly Progress Reports 1957; various authors; SRDB 

Ref ID: 85474 

HP Monthly Reports, 1959-1960, Health Physics Monthly Reports 1959-1960; J. W. McCaslin; 

SRDB Ref ID: 138195, PDF pp. 37-38, 86-87, 105-106, 123, 152 

HP Monthly Reports, 1962, Health Physics Monthly Reports - 1962, Phillips Petroleum; various 

authors; monthly for 1962; SRDB Ref ID: 138197 

HP Regs, 1963, Health Physics Regulations Related to L Cell and RaLa Processing, Standard 

Practices Manual, 6.10.3; author unknown; March 14, 1963; SRDB Ref ID: 138258 

Hueter, 1963, Radioactive Waste Reports, MTR Liquid and Airborne and ETR Liquid, OctoberNovember 1963: L. J. Hueter; 1963; SRDB Ref ID: 138892 

Hughes, 2002, Building 707 Facility Complex Decommissioning Basis for Interim Operation (DBIO) 

Revision 4; S. R. Hughes; February 2002; SRDB Ref ID: 104875 

I-131 Containment, 1961, Containment of Iodine-131 Released by the RaLa Process, IDO-14566, G. 

K. Cederberg; Phillips Petroleum Company; October 30, 1961; SRDB Ref ID: 133328 

ICPP HP Upgrade, 1971-1980, Preliminary ICPP Health Physics Upgrade Program; various authors, 

1971-1980; SRDB Ref ID: 124516 

IDO Report, 1961, IDO Report on the Nuclear Incident at the SL-1 Reactor, IDO-19302; author 

unknown; January 1962; SRDB Ref ID: 87848 

Incoming Radioactive Shipment Records, 1961, Incoming Radioactive Shipments April – December 

1961; author unknown; SRDB Ref ID: 140188 
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INEEL, 1998, Historic American Engineering Record – Photographs, Written Historical and 

Descriptive Data for Various Years; Idaho National Engineering and Environmental Laboratory 

(INEEL); Received September 15, 1998; SRDB Ref ID: 139434; SRDB Ref ID: 118550 

INEEL, 2004, Removal Action Work Plan for the Decontamination and Decommissioning of Building 

CPP-627, the Remote Analytical Facility, DOE/NE-ID-11158; Idaho National Engineering and 

Environmental Laboratory (INEEL); June 2004; SRDB Ref ID: 139435 

INEEL, 2005, Historical American Engineering Record—Idaho National Engineering and 

Environmental Laboratory—Test Area North, HAER No. ID-33-E, INEEL/EXT-04-02536; Idaho 

National Engineering and Environmental Laboratory (INEEL); February 2005; SRDB Ref ID: 83751 

INEL Bioassay, 2010a, INEL Bioassay Records, CD 1 of 3, Idaho National Laboratory; historical 

records (10,583 pages) collected by NIOSH on September 1, 2010; various dates and years; SRDB 

Ref ID: 123095 

INEL Bioassay, 2010b, INEL Bioassay Records, CD 2of 3, Idaho National Laboratory; historical 

records (10,583 pages) collected by NIOSH on September 1, 2010; various dates and years; SRDB 

Ref ID: 123095 

INL, 1963, Brief SNAPTRAN Program Summary; Idaho National Laboratory (INL); Circa 1963; 

SRDB Ref ID: 139088

INL, 2006; Final Report for the Decontamination and Decommissioning of CPP-627 Remote 

Analytical Facility, ICP/EXT-06-01145; Idaho National Laboratory (INL); February 2006; SRDB Ref 

ID: 139433 

INL, 2009, Idaho National Laboratory Cultural Resource Management Plan, Rev. 3, DOE/ID-10997; 

Idaho National Laboratory (INL); February 2009; SRDB Ref ID: 119133 

Irvine, 1954, Idaho Chemical Processing Plant Process and Laboratory Building Operating Manual, 

Volume 1, IDO-14277; V. W. Irvine; June 1954; SRDB Ref ID: 87771 ECI

Johnson, 1961, MTR Health Physics Report for July 1961; Joh-25-61A; Blair D. Johnson; August 18, 

1961; SRDB Ref ID: 138196, PDF pp. 12, 180-181 

Johnson, 1963, MTR Health Physics Report for November, 1963; Joh-3-64A; Blair D. Johnson; 

January 8, 1964; SRDB Ref ID: 138880, PDF pp. 5-6 

Johnson, 1965, MTR Health Physics Report for December, 1964, LJJ-2-65A; LaMar J. Johnson; 

January 14, 1965; SRDB Ref ID: 139193, PDF p. 3 

Johnson, 1965a, TRA Hot Cell Contamination 8-9-65; LJJ-36-65A; L. J. Johnson; August 10, 1965; 

SRDB Ref ID: 125970, PDF pp. 84, 116, 122-125, 131, 133 

Johnson, 1969, Pu and Am-241 Levels in Rocky Flats Wastes at INL; October 13, 1969; Blair 

Johnson; SRDB Ref ID: 125546 
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Johnston, 1952, Health Physics Services Provided at the NRTS, ID-B&M; L. E. Johnston; April 14, 

1952; SRDB Ref ID: 139103 

Kalbeitzer, 1983, Reportable Threshold Reading from the Unshielded Dosimeter of the INEL 

Personnel Badge, memorandum to Dosimetry Branch, F. L., Kalbeitzer; Radiological and 

Environmental Sciences Laboratory; Idaho Falls, Idaho; March 25, 1983; SRDB Ref ID: 8154 

Keller, 1958, Air Activity Incident – MTR Reactor Building, December 15, 1958, correspondence to R. 

L. Doan; F. R. Keller; December 16, 1958; SRDB Ref ID: 139186 

Keller, 1958a, MTR Progress Report Cycles No. 108 and 109, August 4 thru September 8, 1958, IDO16496; F. R. Keller; October 1, 1958; SRDB Ref ID: 83102 

Kunze, 1970, Facility Operating Manual-Low Power Test Facility, ANCR-1137; J. F. Kunze and P. 

L. Chase; December 1970; SRDB Ref ID: 1037 

Latchum, 1963, Findings of Accident Review Committee, Alpha Contamination Incident, Test Reactor 

Area (TRA) November 15-18, 1963; J. W. Latchum; December 10, 1963; SRDB Ref ID 88068, PDF 

pp. 1-62 

Layfield, 1961, SL-1 Health Physics Data Report No. III, DR-RO-3; R. L. Layfield; February 15, 

1961; SRDB Ref ID: 83775 

Layfield, 1961a, SL-1 Health Physics Data Report No. IV, DR-RO-4; R. L. Layfield; February 22, 

1961; SRDB Ref ID: 83653 

Legler, 1957, Startup Operation of a Production Facility for Separating Barium-140 from MTR Fuel, 

IDO-14414; B. M. Legler, S. F. Fairbourne, P. N. Kelly, and R. A. Robinson; September 1, 1957; 

SRDB Ref ID: 136831

Levine, 1959, General Operating Rules for Nuclear Testing at the NRTS – ANP Site, DC-59-4-712; S. 

J. Levine; April 3, 1959; SRDB Ref ID: 116980 

LOC, 1957-2014, Idaho National Laboratory Photos; compilation of links to various photos accessed 

through the Library of Congress (LOC); SRDB Ref ID: 138147 

Long, 1969, Fast Burst Reactors, R. L. Long and P. D. O’Brien; U.S. Atomic Energy Commission, 

Division of Technical Information; Proceedings of the National Topical Meeting on Fast Burst 

Reactors; University of New Mexico, January 28-30, 1969; SRDB Ref ID: 136474

Loop, 1963, Plutonium Contamination at the Materials Testing Reactor, Conclusions and 

Recommendations of the Investigating Committee; E.K. Loop, November 18, 1963; SRDB Ref ID 

88068, PDF pp.17, 63-72 

Loop, 1963a, Plutonium Contamination at the Materials Testing Reactor, Findings of the 

Investigating Committee; E. K. Loop, November 18, 1963; SRDB Ref ID: 88068, PDF pp. 40-41 
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Mahathy, 2015, Mike Mahathy’s OSTI Notes of Gas Analysis Highly Irradiated Be Sample; Mike 

Mahathy; 2015; SRDB Ref ID: 141280 

Malmstrom, 1972, Plutonium and Uranium Inhalation Exposures at the Analytical Chemistry 

Laboratory of the Idaho Chemical Processing Plant; Carl R. Malmstrom; December 4, 1972; SRDB 

Ref ID 116077 

Mandiloff, 1971, Phoenix Core Fuel Incident Report, Man-3-71; D. M. Mandiloff; January 20, 1971; 

SRDB Ref ID: 140929

McCaslin, 1953, Radioactive Shipment Monitoring Records January – June 1953; J. W. McCaslin; 

SRDB Ref ID: 138870 

McCaslin, 1963, Whole Body Counting Procedures Standard Practices Manual; 6.24.1; J. W. 

McCaslin; April 1, 1963; SRDB Ref ID: 132561 

McCaslin, 1966, J. W. McCaslin Correspondence 1966; J. W. McCaslin; SRDB Ref ID: 126064 

McCaslin, 1966a, J. W. McCaslin Correspondence 1966; J. W. McCaslin; SRDB Ref ID: 126064, 

PDF p. 19 

McKenna, 1963, Radioactive Material Shipment Record; J. H. McKenna; April 26, 1963; SRDB Ref 

ID: 140429, PDF p. 19 

Measurement Procedures, 1983, Radiologic Field Measurement Procedures; GJ-09(83); author 

unknown; July 1983; SRDB Ref ID: 94140 

Mills, 1957, SPERT Incident Report 3-18-57 – 3-22-57, RLM-8-57A, R. L. Mills; March 29, 1957; 

SRDB Ref ID: 86640, PDF pp. 2-3 

Mills, 1957a, Multicurie Cell Facility Decontamination; McV-10-56A; W. H. McVey; August 24, 

1956; SRDB Ref ID: 86640, PDF p. 60 

Moser, 1985, RALA Off-Gas Cell and Storage Tank (CPP-631 and VES-702) Decontamination and 

Decommissioning Plan, WINCO-1032; C. L. Moser, Westinghouse Idaho Nuclear Company, Inc.; 

July 1985; SRDB Ref ID: 83550 

MTR, 1960s, Materials Testing Reactor, Idaho Nuclear Corporation; date unknown, but from context: 

mid-1960s; SRDB Ref ID: 139250 

MTR, 1967, MTR Phoenix Core; C. W. Bills, J. R. Horan, E. K. Loop; 1967; SRDB Ref ID: 116086 

MTR Progress Report, 1953, Materials Testing reactor Progress Report: December 1953, Phillips 

Petroleum Company; issued January 8, 1954; SRDB Ref ID: 83033, PDF p. 24 

Murphy, 1966, Final Disassembly and Examination of the ML-1 Reactor Core, IDO-17190; T. L. 

Murphy, D. R. Schuyler, and W. L. Clarke; issued June 1966; SRDB Ref ID: 141102 
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Nelson, 1961, Report on the SL-1 Incident, January 3, 1961; author unknown; May 1961; SRDB Ref 

ID: 4495 

Neptunium, 1969, Neptunium Purification at ICPP, An-267-69, F. H. Anderson; Idaho Nuclear 

Corporation; October 9, 1969; SRDB Ref ID: 136643 

Neptunium, 1972, Neptunium-237: Production and Recovery, ARH-2417, W. W. Schultz; Atlantic 

Richfield Hanford Company; March 21, 1972; SRDB Ref ID: 105831 

Neptunium, 1973, First Neptunium Processing Campaign at ICPP, ICP-1019, J. E. Johnson; Allied 

Chemical Corporation; February 1973; SRDB Ref ID: 130210 

Newby, 2000, Summary of Waste Calcination at the Idaho Nuclear Technology and Engineering 

Center, INEEL/EXT-2000-01206; B. J. Newby and B. H. O’Brien; October 2000; SRDB Ref ID: 

88549 

Notes, 2015, [Name’s redacted] Notes on TRA Progress Reports, [Name redacted], handwritten notes 

made during review of TRA Progress Reports regarding various TRA facilities; January 8, 2015; 

SRDB Ref ID: 140873

NRTS Facilities, 1973, Tabulation of Facilities at the National Reactor Testing Station, author 

unknown; National Reactor Testing Station; July 1973: SRDB Ref ID: 29250 

Offutt, 1969, Rare Gas Recovery Facility at the Idaho Chemical Processing Plant, IN-1221; G. F. 

Offutt and C. L. Bendixsen; April 1969; SRDB Ref ID: 136144 

OMRE, 1957, OMRE – Another U.S. Project Under Way, author unknown; Nuclear Engineering, 

December 1957; SRDB Ref ID: 139241 

OMRE, 1959, OMRE Radioactive Waste Reports 1959; author unknown; 1959; SRDB Ref ID: 87001 

OMRE, 1961, Radioisotopic Investigating of OMR Coolants, Part I: OMRE Operation January, 1958 

Through August, 1960; NAA-SR-5049, L. Baurmash; Atomics International; December 1, 1961; 

SRDB Ref ID: 85399

OMRE HP Survey, 1962-1963, Organic Moderated Reactor Experiment Health Protection Surveys, 

1962-1963; SRDB Ref ID: 71410 

ORAUT-OTIB-0005, Internal Dosimetry Organ, External Dosimetry Organ, and IREP Model 

Selection by ICD-9 Code, Rev 05; ORAU Team Dose Reconstruction Project for NIOSH; December 

20, 2012; SRDB Ref ID: 121336 

ORAUT-OTIB-0006, Dose Reconstruction from Occupational Medical X-Ray Procedures, Rev. 04; 

ORAU Team Dose Reconstruction Project for NIOSH; June 20, 2011; SRDB Ref ID: 98147

ORAUT-OTIB-0011, Technical Information Bulletin: Tritium Calculated and Missed Dose Estimates, 

Rev 00; ORAU Team Dose Reconstruction Project for NIOSH; June 29, 2004; SRDB Ref ID: 19430 
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ORAUT-OTIB-0014, Assignment of Environmental Internal Doses for Employees Not Exposed to 

Airborne Radionuclides in the Workplace, Rev 00; ORAU Team Dose Reconstruction Project for 

NIOSH; June 22, 2004; SRDB Ref ID: 19432 

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

Rev 01; ORAU Team Dose Reconstruction Project for NIOSH; October 11, 2005; SRDB Ref ID: 

19434 

ORAUT-OTIB-0018, Internal Dose Overestimates for Facilities with Air Sampling Programs, Rev 

01; ORAU Team Dose Reconstruction Project for NIOSH; August 9, 2005; SRDB Ref ID: 19436 

ORAUT-OTIB-0023, Technical Information Bulletin: Assignment of Missed Neutron Doses Based on 

Dosimeter Records, Rev 01; ORAU Team Dose Reconstruction Project for NIOSH; May 14, 2008; 

SRDB Ref ID: 43937

ORAUT-OTIB-0033, Application of Internal Doses Based on Claimant-Favorable Assumptions for 

Processing as Best Estimates, Rev 00; ORAU Team Dose Reconstruction Project for NIOSH; April 

20, 2005; SRDB Ref ID: 19457 

ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly Retained in the Lung, Rev 01 PC-2; 

ORAU Team Dose Reconstruction Project for NIOSH; November 29, 2010; SRDB Ref ID: 90666 

ORAUT-OTIB-0054, Technical Information Bulletin: Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses, Rev 02; ORAU Team Dose 

Reconstruction Project for NIOSH; March 6, 2014; SRDB Ref ID: 130852 

ORAUT-OTIB-0060, Internal Dose Reconstruction, Rev 01; ORAU Team Dose Reconstruction 

Project for NIOSH; September 8, 2014; SRDB Ref ID: 135808 

ORAUT-OTIB-0079, Guidance on Assigning Occupational X-Ray Dose under EEOICPA for X-Rays 

Administered Off Site, Rev 00; ORAU Team Dose Reconstruction Project for NIOSH; January 3, 

2011; SRDB Ref ID: 89563 

ORAUT-OTIB-0082, Technical Information Bulletin: Dose Reconstruction Method for Chronic 

Lymphocytic Leukemia, Rev 00 PC-1; ORAU Team Dose Reconstruction Project for NIOSH; 

December 20, 2012; SRDB Ref ID: 121337 

ORAUT-PROC-0060, Procedure: Occupational On-Site Ambient Dose Reconstruction for DOE Sites, 

Rev 01; ORAU Team Dose Reconstruction Project for NIOSH; June 28, 2006; SRDB Ref ID: 29986 

ORAUT-PROC-0061, Procedure: Occupational Medical X-Ray Dose Reconstruction for DOE Sites, 

Rev 03; ORAU Team Dose Reconstruction Project for NIOSH; March 3, 2010; SRDB Ref ID: 79758 

ORAUT-TKBS-0007-1, Idaho National Laboratory and Argonne National Laboratory-West – 

Introduction, Rev. 03; ORAU Team Dose Reconstruction Project for NIOSH; Effective Date: March 

12, 2010; SRDB Ref ID: 79862 
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ORAUT-TKBS-0007-2, Idaho National Laboratory and Argonne National Laboratory-West – Site 

Description, Rev. 04; ORAU Team Dose Reconstruction Project for NIOSH; Effective Date: August 

2, 2010; SRDB Ref ID: 84106

ORAUT-TKBS-0007-3, Idaho National Laboratory and Argonne National Laboratory-West – 

Occupational Medical Dose, Rev. 02; ORAU Team Dose Reconstruction Project for NIOSH; 

Effective Date: December 21, 2009; SRDB Ref ID: 77829 

ORAUT-TKBS-0007-4, Idaho National Laboratory and Argonne National Laboratory-West – 

Occupational Environmental Dose, Rev. 02; ORAU Team Dose Reconstruction Project for NIOSH; 

Effective Date: January 8, 2010; SRDB Ref ID: 78635 

ORAUT-TKBS-0007-5, Idaho National Laboratory and Argonne National Laboratory-West – 

Occupational Internal Dose, Rev. 03; ORAU Team Dose Reconstruction Project for NIOSH; 

Effective Date: March 2, 2010; SRDB Ref ID: 79571 

ORAUT-TKBS-0007-6, Idaho National Laboratory and Argonne National Laboratory-West – 

Occupational External Dose, Rev. 03; ORAU Team Dose Reconstruction Project for NIOSH; 

Effective Date: April 19, 2011; SRDB Ref ID: 94104 

Paulus, undated, Report of Investigation, Criticality Incident at the Idaho Chemical Processing Plant; 

R. C. Paulus; undated; SRDB Ref ID: 88441 

Perry, 1993, Minimum Reporting Level for the Idaho National Engineering Laboratory (INEL) 

Thermoluminescent Dosimetry (TLD) Systems, EGG-ESQ-10917; O. R. Perry, B. L. Andersen, and P. 

E. Ruhter; Idaho National Engineering Laboratory; SRDB Ref ID: 8161 

Personal Communication, 2014a, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by NIOSH and SC&A; November 2014; SRDB Ref ID: Not yet in 

SRDB

Personal Communication, 2014b, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, ABRWH, SC&A, and NIOSH; November 17, 2014; 

SRDB Ref ID: Not yet in SRDB

Personal Communication, 2014c, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, ABRWH, SC&A, and NIOSH; November 17, 2014; 

SRDB Ref ID: Not yet in SRDB

Personal Communication, 2014d, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, Board, SC&A, and NIOSH; November 18, 2014; 

SRDB Ref ID: Not yet in SRDB

Personal Communication, 2014e, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, Board, SC&A, and NIOSH; November 18, 2014; 

SRDB Ref ID: Not yet in SRDB
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Personal Communication, 2014f, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by SC&A and NIOSH; November 18, 2014; SRDB Ref ID: Not yet in 

SRDB

Personal Communication, 2014g, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, Board, SC&A, and NIOSH; November 18, 2014; 

SRDB Ref ID: Not yet in SRDB

Personal Communication, 2014h, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team, SC&A, and NIOSH; November 20, 2014; SRDB Ref 

ID: Not yet in SRDB

Personal Communication, 2014i, Personal Communication with former INL worker; In-person 

interview at Idaho Falls, Idaho by ORAU Team and NIOSH; December 11, 2014; SRDB Ref ID: Not 

yet in SRDB

Peterson, 2004, Supplement to Technical Basis Document 4 for the Idaho National Engineering and 

Environmental Laboratory: INEEL Occupational Environmental Dose, Rev. 0; Henry Peterson; 

March 9, 2004; SRDB Ref ID: 15451

Petition, 2014, Special Exposure Cohort Petition - Form B for Idaho National Laboratory, received 

by NIOSH on July 18, 2014; DSA Ref ID: 120082 

Phillips, 1962, Phillips Petroleum Company, Atomic Energy Division, Idaho Falls, Idaho, Operation 

overview brochure, Phillips Petroleum Company; from context: about 1962; SRDB Ref ID: 139280 

Phoenix, 1967, Significant Unplanned Event; William C. Phoenix; August 1967; SRDB Ref ID 

138999 

Picker, 2001, TAN Hot Shop and Support Facility Utilization Study, INEEL/EXT-01-01481; B. A. 

Picker; November 2001; SRDB Ref ID: 136250 

Plutonium, 1980, A Review of the Plutonium Problem in ICPP Facilities and Projects, M. J. 

Szulinski; Science Applications, Inc.; April 1980; SRDB Ref ID: 8431 

Plutonium in Feces, 1965, Determination of Plutonium in Feces; author unknown; July 7, 1965; 

SRDB Ref ID: 140462

Progress Report, 1968, Health and Safety Branch Progress Report – 1968, J. W. McCaslin; Idaho 

Nuclear Corporation; 1968; SRDB Ref ID: 60289 

Progress Reports, 1959, Health Physics Monthly Progress Reports for 1959; various authors; various 

dates throughout 1959; SRDB Ref ID: 85456 

Puphal, 1994, History of the INEL Bioassay Programs; Puphal; May 2, 1994; SRDB Ref ID: 14664 
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Redesignation, 1963, Redesignation of the Army Reactor Experiment Area at the NRTS, H. N. 

Eskildson; U. S. Atomic Energy Commission, Idaho Operations Office; July 12, 1963; SRDB Ref ID: 

138181 

Remediation, 2006, Remedial Investigating and Baseline Risk Assessment for Operable Unit 7 13/14, 

K. J. Holdren; Idaho Cleanup Project; U.S. Department of Energy, Idaho Operations Office; May 

2006; SRDB Ref ID: 138535 

RFP Waste, 2005, Historical Background Report for Rocky Flats Plant Waste Shipped to the INEEL 

and Buried in the SDA from 1954 to 1971, ICP/EXT-04-00248; North Wind, Inc., TRU Program 

Technical Services; February 2005; SRDB Ref ID: 83756 

Rich, 1958, Thyroid Exposures from March 1958 Incident at ICPP; BLR-83-58A; Bryce L. Rich; 

April 21, 1958; SRDB Ref ID: 8013 

Rich, 1974, Preliminary ICPP Health Physics Upgrade Program; ACI-167; B. L.Rich; October 1974; 

SRDB Ref ID: 14667 

Rich, 1974a, CPP-603 Contamination Control Status, BLR-6-74; B. L. Rich; March 1, 1974; SRDB 

Ref ID: 118914 

Rich, 1979, 1978 Health Physics Annual and Upgrade Status Report, BLR-10-79; B. L. Rich; May 3, 

1979; SRDB Ref ID: 124517 

Robertson, 1982, Brief History of GE at INL and Disposition of Records; D. C. Robertson; March 26, 

1982; SRDB Ref ID: 90564 

Ruhter, 2002, Idaho National Engineering and Environmental Laboratory (INEEL) Dosimetry 

Systems History and Minimum Dose Reporting Level (MRL) Information (External Dose Summary 

Profile – 1951 to March 2002), P. E. Ruhter and O. R. Perry; Bechtel BWXT Idaho, Idaho National 

Engineering and Environmental Laboratory; April 8, 2002; SRDB Ref ID: 1774, PDF p. 2 

RWMC History, 1979, A History of the Radioactive Waste Management Complex at the Idaho 

National Engineering Laboratory; PR-W-79-038; B. C. Anderson; Operations Project, Waste 

Management Programs, EG&G Idaho, Inc.; SRDB Ref ID: 83011

RWMC History, 1985, A History of the Radioactive Waste Management Complex at the Idaho 

National Engineering Laboratory; WM-F1-81-003, Rev. 3; author unknown; Waste Program 

Division, RWMC Operations Branch; July 1985; SRDB Ref ID: 83913; SRDB Ref ID: 124478 

TAN Shipments, 1963-64, Radioactive Material Shipment Records for TAN, 1963-64, various authors 

and dates; Idaho National Laboratory; SRDB Ref ID: 138229 

SABER, 1990, Steam Accumulator Blowdown Evaluation Rig Large Scale Steam Valve Test Action 

Description Memorandum; author unknown; SRDB Ref ID: 139732 
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Savignac, 1961, CFA Health Physics Monthly Report for September, 1961; Sav-12-61A; Bart 

Savignac; October 24, 1961; SRDB Ref ID: 139258 

Savignac, 1961a, CFA Health Physics Monthly Report for July, 1961; Sav-9-61A; Bart Savignac; 

August 11, 1961; SRDB Ref ID: 139258, PDF p. 13 

Savignac, Jul1961, CFA Health Physics Monthly Report for July, 1961, Sav-9-61A; Bart Savignac; 

August 11, 1961; SRDB Ref ID: 138196 

Savignac, 1963, Monthly Burial Ground Report; Sav-11-63A; B. F. Savignac; May 8, 1963; SRDB 

Ref ID: 139220 

Sauvignac, 1963a, Monthly Burial Ground Report October – December 1962; Bart Sauvignac; 1962-

1963; SRDB Ref ID: 138684 

Shank, 1964, Annual Report of Division Analytical Branch for 1963; IDO-14636; Ralph C. Shank; 

June 1964; SRDB Ref ID: 138063 

Shank, 1969, Analytical Chemistry Branch Annual Report Fiscal Year 1969; IN-1316; R. C. Shank; 

September 1969; SRDB Ref ID: 136032 

Shipment Records, 1960s, Radioactive Shipment Records for U-235, Mixed Fission Products and 

Contaminated Respirators, U.S. Atomic Energy Commission, Idaho Operations Office; various 

authors; various dates in early 1960s; SRDB Ref ID: 139190 

SL-1 Chronology, 1961, SL-1 Chronology, January through April 1961, author unknown; Idoho 

National Laboratory; February 2, 1961; SRDB Ref ID: 83755 

SL-1 Final, 1962; Final Report of SL-1 Recovery Operation, IDO-19311; author unknown, July 27, 

1962; SRDB Ref ID: 140097 

SL-1 IDO Report, 1962, IDO Report on the Nuclear Incident at the SL-1 Reactor January 3, 1961 at 

the National Reactor Testing Station; IDO-19302; author unknown; January 1962: SRDB Ref ID: 

85418 

SL-1 Records, 2009, Historical Information for SL-1 (Accident Records), SL-1 accident response 

timeline, observations, log entries, badge reports, dosimetry reports and misc. documents gathered in 

2009 by ORAU Team; SRDB Ref ID: 59343 

SL-1 Report, 1961, Report on the SL-1 Incident, January 3, 1961, C. A. Nelson; U.S. Atomic Energy 

Commission Board of Investigation; May 1961; SRDB Ref ID: 4495 

SL-1 Safety, 3Q-1959, ABWR Quarterly Progress Report, Volume II, SL-1 Health Physics & Safety, 

July 1 to September 30, 1959, IDO-19005, E. J. Vallario; Combustion Engineering; October 15, 1959; 

SRDB Ref ID: 8142
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SL-1 Safety, 4Q-1959, ABWR Quarterly Progress Report, Volume II, SL-1 Health Physics & Safety, 

October 1 to December 31, 1959, IDO-19009, E. J. Vallario; Combustion Engineering; January 15, 

1960; SRDB Ref ID: 140236 

SL-1 Safety, 1Q-1960, ABWR Quarterly Progress Report, SL-1 Health Physics & Safety, January 1 to 

March 31, 1960, IDO-19013, R. T. Canfield; Combustion Engineering; May 25, 1960; SRDB Ref ID: 

135863 

SL-1 Safety, 2Q-1960, ABWR Quarterly Progress Report, SL-1 Health Physics & Safety, April 1 to 

June 30, 1960, IDO-19017, W. B. Allred; Combustion Engineering; August 12, 1960; SRDB Ref ID: 

102428 

SL-1 Sr-90 Dose, 1961, Probable Strontium-90 Dose of Personnel Involved with SL-1 Incident, 

January 3, 1961, E. R. Ebersole; Analysis Branch, Health and Safety Division, Idaho National 

Laboratory; document date unknown; SRDB Ref ID: 31960 

Smith, 1981, Characterization of the RALA Off-Gas Cell, CPP-631, WM-F1-81-010; D. L. Smith and 

D. J. Bradford; May 1981; SRDB Ref ID: 31990 

SNAPTRAN, 1964, SNAPTRAN 2/10A-3 Program, Doc. No. 726881, Idaho National Laboratory; 

May 1964; SRDB Ref ID: 136013 

SNAPTRAN, 1967, SNAPTRAN-2 Destructive Test Results, IDO-17194, W. E. Kessler; Phillips 

Petroleum Company; January 1967; SRDB Ref ID: 140002 

Sommers, 1956, Radioactivity Incident Reports, John F. Sommers, January 1956; SRDB Ref ID: 

139301, PDF pp. 60-69 

Sommers, 1957, Crystal Spectrometer Incident – 12-11-57; Som-122-57A; John F. Sommers; 

December 13, 1957; SRDB Ref ID: 85448, PDF pp. 104-105 

Sommers, 1957a, Hot Cell Contamination Incident on March 8, 1957; Som-34-57A; John Sommers; 

March 29, 1957; SRDB Ref ID: 86568, PDF pp. 18-20, 105-108 

Sommers, 1959, Radioactive Shipment Record Details of Shipment MTR Hot Laundry December 

1959; John Sommers; SRDB Ref ID: 138144 

Sommers, 1959a, MTR Health Physics Program Report for August, 1959; Som-80-59A; John F. 

Sommers; September 16, 1959; SRDB Ref ID: 87081, PDF p. 23 

Sommers, 1961, Whole Body Counting Procedures for MTR-ETR HP’s; Som-106-61A-M; John F. 

Sommers; September 27, 1961; SRDB Ref ID: 8545 

Sommers, 1963, Plutonium Contamination Incident at the TRA November 13-18, 1963; John F. 

Sommers; undated; SRDB Ref ID: 87466 
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Sommers, 1966, Whole Body Counting; Som-44-66; John F. Sommers; September 6, 1966; SRDB Ref 

ID: 32048 

Sommers, 1968, TRA Routine Whole Body Counts in 1968; Som-5-68; John F. Sommers; January 11, 

1968; SRDB Ref ID: 32054 

Special HP Services, 1968, Idaho Nuclear Corporation Standard Practice No. 6.20.2, Special HP 

Services, F. A. Anderson; Idaho Nuclear Corporation; October 17, 1968; SRDB Ref ID: 141098 

SPERT, 1957, Various SPERT Activity Release Reports for 1957; Phillips Petroleum Company; 

various dates throughout 1957; SRDB Ref ID: 31877 

SPERT and SNAPTRAN, 1960s, Source Term Quality Control Tasks (Episodic Releases), DOE/ID12119-QAF-014 6- 018, assorted documents from 1991 focused on SPERT events in the 1960s; also 

includes SNAPTRAN events; various authors; Idaho National Laboratory; various dates; SRDB Ref 

ID: 88727 

SPERT-I, 1961, Special Power Excursion Reactor Tests (SPERT), General Information, Fact Sheet 

on SPERT-I, Phillips Petroleum Company; April 2, 1961; SRDB Ref ID: 136955 

SPERT-I, 1965, SPERT I Destructive Test Series Environmental Monitoring and Research Studies, 

IDO-12039, D. F. Bunch; Idaho Operations Office, U.S. Atomic Energy Commission; January 1965; 

SRDB Ref ID: 83580

SPERT-I, -II, and –III, 1957, Summary of the SPERT-I, -II, and –III Reactor Facilities, C. R. 

Montgomery; Phillips Petroleum Co.; November 1, 1957; SRDB Ref ID: 136833 

SPERT-III, 1961, SPERT III Reactor Facility, R. E. Heffner; National Reactor Testing Station, U.S. 

Atomic Energy Commission; October 25, 1961; SRDB Ref ID: 139856 

SPERT-IV, 1962, SPERT IV Facility, IDO-16745, R. E. Heffner, Phillips Petroleum Company; 

February 22, 1962; SRDB Ref ID: 139885 

SPERT Dosimetry, 1957-58, Various SPERT Documents Regarding Dosimetry, various authors and 

dates from January 2, 1957 to October 7, 1958; Idaho National Laboratory; SRDB Ref ID: 140361 

SPERT Program, 1957, SPERT Program Review, W. E. Nyer; Phillips Petroleum Co.; September 27, 

1957; SRDB Ref ID: 136953 

SPERT Tests, 1965, Special Power Excursion Reactor Tests, Phillips Petroleum Company; August 

13, 1965; SRDB Ref ID: 116091 

Stacy,1995, Historical American Engineering Record, Idaho National Engineering Laboratory, Test 

Area North, Hangar 629, HAER No. ID-33-A; Susan Stacy; August 12, 1995; SRDB Ref ID: 136966 

Stacy, 2000, Proving the Principle, DOE/ID-10799; Susan M. Stacy, Idaho Operations Office of the 

Department of Energy; Published in 2000; SRDB Ref ID: 1152 
SEC-00219 03-12-15 Idaho National Laboratory

242 of 258 

Stacy, 2000a, Proving the Principle, DOE/ID-10799; Susan M. Stacy, Idaho Operations Office of the 

Department of Energy; Published in 2000; SRDB Ref ID: 1152, PDF p. 279

Stacy, 2000b, Proving the Principle, DOE/ID-10799; Susan M. Stacy, Idaho Operations Office of the 

Department of Energy; Published in 2000; SRDB Ref ID: 1152, PDF p. 274

Stacy, 2006, Historic American Engineering Record, Idaho National Laboratory, Idaho Chemical 

Processing Plant, Fuel Reprocessing Complex, INL/EXT-06-11969; Susan Stacy and Julie Braun; 

December 2006; SRDB Ref ID: 80142 

Stroschein, 1967, Health Physics Manual of NRTS Areas Operated by Idaho Nuclear Corporation; 

edited by H. W. Stroschein and P. H. Maeser; Issued: April 1967; SRDB Ref ID: 1029 

Survey, Aug-Dec1953, Alpha Smear Surveys (August-December 1953); August 31-December 16, 

1953; SRDB Ref ID: 138873 

Survey, May-Jun1955, Alpha Smear Surveys (May-June 1955); May 26-June 30, 1955; SRDB Ref ID: 

138871 

Survey, Oct1955, Alpha Smear Surveys for Main Floor, Reactor Basement, Labs 106, 107, 109, 112, 

and 113 (October 1958); October 9-25, 1955; SRDB Ref ID: 138724 

Survey, Jun1958, Alpha Beta Gamma Surveys (June 1958); June 1-30, 1958; SRDB Ref ID: 138681 

Survey, Nov1958, Alpha Surveys (November 1958); November 1-30, 1958; SRDB Ref ID: 138688 

Survey, Apr1963, MTR Health Physics Survey Logs (April 1963); April 1-30, 1963; SRDB Ref ID: 

140256 

Survey, Jul1963, HP Logbook Sheets MTR with Total Daily Swipe Counts and Beta and Gamma 

Surveys (July 1963); July 1-31, 1963; SRDB Ref ID: 140359 

Survey, Aug1968, MTR Contamination Surveys (August 1968); August 1-31, 1968; SRDB Ref ID: 

140292 

Survey, Apr1969, MTR HP Survey Logs (April 1969); March 31-April 30, 1969; SRDB Ref ID: 

140315 

Surveys, 1968, Floor Plan Surveys July 1968; author unknown; 1968; SRDB Ref ID: 138812 

SWP 1962-1964, Safe Work Permit Burial Ground January 1962 – December 1964; author unknown; 

1962-1964; SRDB Ref ID: 138194 

TAN Alpha Survey, 1968, Test Area North Alpha Survey, Hot Cell Area; W. Anderson; June 4, 1968; 

SRDB Ref ID: 140265, PDF p. 119 
SEC-00219 03-12-15 Idaho National Laboratory

243 of 258 

TAN HP Report, Dec1963, TAN Health Physics Report for December, 1963, O. L. Cordes; Phillips 

Petroleum Company, Atomic Energy Division, Idaho Falls, Idaho; January 9, 1964: SRDB Ref ID: 

138714 

TAN Monitoring, 1963, Tan Area Monitoring Schedule; author unknown; March 1963; SRDB Ref 

ID: 140265, PDF p. 105 

Terch, 1963, SPERT Health Physics Monthly Report for August, 1963, Ter-10-63A; L. P. Terch; 

September 10, 1963; SRDB Ref ID: 138710 

Thorium, 1961, Remote Dissolution and Analytical Program for Irradiated Thorium, IDO-14557, G. 

A. Huff; Phillips Petroleum Company; July 14, 1961; SRDB Ref ID: 133336 

Thornton, Feb1962, Comprehensive Technical Report – General Electric Direct-Air-Cycle Aircraft 

Nuclear Propulsion Program, Heat Transfer Reactor Experiment No. 1, APEX-904; G. Thornton, S. 

H. Minnich, C. Heddleson, and D. H. Culver; February 28, 1962; SRDB Ref ID: 84049 

Thornton, Jun1962, Comprehensive Technical Report – General Electric Direct-Air-Cycle Aircraft 

Nuclear Propulsion Program, Program Summary and References, APEX-901; G. Thornton and A. J. 

Rothstein; June 28, 1962; SRDB Ref ID: 1078 

Tiger Team, 1991, Tiger Team Assessment of the Idaho National Engineering Laboratory Volume 2, 

Section 4.0; author unknown; 1991; SRDB Ref ID: 77146 

Tingey, 1968, Qualification of Umpire Laboratories for the Analysis of Uranium and Plutonium 

Materials Specifications Report 1, Basic Programs; F. H. Tingey; May 1968; SRDB Ref ID: 139246 

Uranium, 1957, Determination of Uranium in Irradiated Thorium, IDO-14411, B. E. Paige; Phillips 

Petroleum Co.; June 19, 1957; SRDB Ref ID: 139893 

Uranium, 1966, Uranium Recovery from Aluminum Alloyed Fuel, ICPP Run No. 21, IDO-14669, B. 

R. Wheeler; Phillips Petroleum Company; January 1966; SRDB Ref ID: 133346 

Voelz, 1969, 1968 Annual Report of the Health Services Laboratory, IDO-12071; George L. Voelz 

and multiple other authors; April 1969; SRDB Ref ID: 1359 

Voelz, 1970, 1969 Annual Report of the Health Services Laboratory, IDO-12073; George L. Voelz 

and multiple other authors; July 1970; SRDB Ref ID: 1362 

Waste Management, 1974, ID Waste Management Plan for FY 1975: National Reactor Testing 

Station; Office of Waste Management, Idaho Operations Office, U.S. Atomic Energy Commission; 

July 1974; SRDB Ref ID: 139837 

Waste Shipments, Feb1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in February 1963; SRDB Ref ID: 139231 
SEC-00219 03-12-15 Idaho National Laboratory

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Waste Shipments, Mar1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in March 1963; SRDB Ref ID: 139232 

Waste Shipments, Apr1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in April 1963; SRDB Ref ID: 139233 

Waste Shipments, May1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in May 1963; SRDB Ref ID: 139234 

Waste Shipments, Jun1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in June 1963; SRDB Ref ID: 139235 

Waste Shipments, Jul1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in July 1963; SRDB Ref ID: 139236 

Waste Shipments, Aug1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in August 1963; SRDB Ref ID: 139237 

Waste Shipments, Nov1963, Disposal of Off-Site Radioactive Waste – Shipment Data Forms, Phillips 

Petroleum Company; various dates in November 1963; SRDB Ref ID: 139238 

WBC1, 1961-1977, INEL Whole Body Count CD 1 of 5; SRDB Ref ID: 123105 

WBC2, 1963-1995, INEL Whole Body Count CD 2 of 5; SRDB Ref ID: 123113 

WBC3, 1981-1994, INEL Whole Body Count CD 3 of 5; SRDB Ref ID: 123114 

WBC4, 1979-1995, INEL Whole Body Count CD 4 of 5; SRDB Ref ID: 123115 

WBC5, 1973-1996, INEL Whole Body Count CD 5 of 5; SRDB Ref ID: 123116 

Wenzel, 1973, Internal Pu Doses; Wen-2-73; D. R. Wenzel; July 24, 1973; SRDB Ref ID: 8014 

Wenzel, 1973b, Internal Doses Resulting from ICPP Analytical Chemistry Laboratory Plutonium 

Contamination Occurrence, correspondence with attached report; D. R. Wenzel; July 24, 1973; SRDB 

Ref ID: 1231 

Willis, 1959, CAM Filter Counting Procedure; CLW-3-59A-N; C. L. Willis; August 19, 1959; SRDB 

Ref ID: 124536, PDF p.50 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Attachment 1: Data Capture Synopsis
 

Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Primary Site/Company Name: Idaho National Laboratory 

(INL) 

DOE 1949 - Present 

Alternate Site Names: 

National Reactor Testing Station 

Idaho National Engineering Laboratory 

Idaho National Engineering and Environmental Laboratory 

Physical Size of the Site: 892 square miles 

Site Population: 4,100 in 2010 

Routine blood sampling, incident reports, environmental reports, SL-1 

reports, internal dosimetry programmatic documents, Aircraft Nuclear 

Propulsion reports, dosimetry report area codes, uranium urinalysis 

using the KPA laser, neutron dose rates, building lists, site telephone 

directories, individual medical records, individual dosimetry reports, 

offsite environmental surveillances, onsite environmental surveillances, 

monitoring and modeling effluent releases, extracting radionuclides 

from product streams, occupational and environmental air sampling 

data, periodic Health Physics reports, periodic facilities reports, tritium 

reports, safety analyses for facilities and campaigns, surveys of specific 

areas with high radionuclide concentrations, ecological reports, 

contamination control plans, logbooks, bioassay result sheets, 

temporary badge reports, correspondence files, Idaho Chemical 

Processing Plant (ICPP) routine surveys, Health Physics log sheets, safe 

work permits, annual exposure records, Health and Safety Record 

Sheets, periodic operations and technical reports, irradiated thorium 

reports, waste handling reports, shipping reports, air filter spectra, and 

analytical methods. NOTE: Awaiting documents to be released from 

INL. 

OPEN 3,449 

State Contacted: NA Since INL is an operating DOE facility the state was not contacted. 01/30/2015 0 

Albany Research Center (ARC) Fluorometric determination of uranium in air dusts and smears, INL 

work performed at ARC, a study on melting wastes, and the disposal of 

solid packaged waste. 

03/20/2013 4 

Argonne National Laboratory (ANL) The 2004 ANL EEOICPA Reasonable Search Procedures, 

decontamination of plutonium contaminated gloveboxes, and monthly 

Operation Clean Sweep status reports from 1982. 

03/26/2008 3 

Battelle Memorial Institute - King Avenue Battelle reactor decommissioning core materials shipped to INL and 

1955-1956 applications to obtain radioisotopes. 

08/18/2014 3 

Brookhaven National Laboratory Ambient air monitoring at DOE facilities, the development of a single 

unit beta-gamma extremity dosimeter, identification of INL as the 

repository for the High Flux Beam Reactor spent fuel, and accelerator 

and medical reactor radiation exposures. 

11/14/2008 5 

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SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Cincinnati Public Library A 1974 history of the Atomic Energy Commission. 02/10/2011 1 

Claimant Provided A petition on the Nuclear Regulatory Commission's oversight of mill 

tailings that refers to INL, the inter-agency reports on links between 

exposure to hazards and illnesses in the DOE contractor workforce, and 

the explosive demolition of the Organic Moderated Reactor Experiment 

building foundation. 

11/26/2013 3 

Colorado Mesa University, Tomlinson Library ICPP waste inventories and properties, removal of radionuclides from 

waste evaporator condensates, and the pilot plant operation of a 

geometrically safe denitrator at the ICPP. 

10/15/2012 3 

Colorado State University A 1986 decontamination and waste treatment facilities assessment. 04/10/2006 1 

Curtiss-Wright, Cheswick, PA The Liquid Metal Fast Breeder Reactor Program, an INL request for 

quotation on plutonium pellets, and shipping information 1968-1972. 

04/24/2009 3 

Dade Moeller External and internal dosimetry records, radiation incident reports and 

dosimetry, air monitoring data, effluent releases, bus survey data, 

surveillances, whole body counts and scheduling, and a records clean 

out campaign. 

04/24/2009 92 

Dade Moeller / SC&A Sample thyroid counts, release reports, incident reports, monthly Health 

Physics reports, and waste reports. 

03/19/2007 11 

DOE Albuquerque Description of INL facility for storing Pantex components. 04/13/2010 1 

DOE Environmental Management Consolidated Business 

Center (EMCBC), Denver 

The 1989 Performance Indicator Pilot Program. 08/15/2013 1 

DOE Germantown A 1958 trip report, Mallinckrodt airport waste, the DOE Legacy 

Management process to respond to EEOICPA claims, search procedures 

for the DOE Oak Ridge Operations Records Holding Area, the 2010 

ANL EEOICPA Reasonable Search Procedures, and a NIOSH 

researcher's notes from a Germantown classified document review. 

03/18/2014 6 

DOE Idaho Operations Office DOELAP external dosimetry reports, survey reports, photographs of 

film packs, film calibrations, dosimeters, Bonner spheres, neutron cards, 

and summary descriptions of film calibration curves and area 

photographs. 

12/20/2013 34 

DOE Legacy Management - Grand Junction Office INL's handling of transuranic waste, a CPAF report, waste 

management, uranium scrap shipments, a 1964 site map with off-site 

environmental monitoring data, bioassay data, programmatic internal 

dosimetry documents, environmental monitoring reports, development 

of reference materials, and an index to reactor fuel data sheets. 

08/25/2011 31 

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Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

DOE Legacy Management - Morgantown Grand Junction Facilities dosimetry reports provided by INL, recycled 

uranium reports, exposure investigations, and an assessment of health 

and mortality studies. 

04/16/2013 105 

DOE Legacy Management - MoundView (Fernald 

Holdings, includes Fernald Legal Database) 

Feed Materials Production Center reports referring to INL, unusual 

occurrence reports, DOE suspends shipments of low-level waste to 

commercial burial grounds, a bioassay conference, a safety analysis 

report for plutonium repackaging, an investigation of a high radiation 

exposure, 1976 Mound transuranic shipments to INL, and the codes for 

the Idaho Operations Office and INL contractors. 

09/01/2008 37 

DOE Legacy Management - Westminster The 2003 analysis of external dosimetry across the DOE Complex, a 

Rocky Flats author list, and INL as a QA laboratory for the Rocky Flats 

bioassay program. 

07/24/2013 3 

DOE Oak Ridge Operations Office Internal audit criteria for industrial hygiene, radiation exposure records 

1954-1968, U-233 disposition strategies and options, and an article on 

radiation injuries. 

07/09/2012 5 

DOE Oak Ridge Operations Office - RHTG An INL shipment to the Portsmouth Gaseous Diffusion Plant and early 

AEC Production Division reports. 

06/25/2013 7 

DOE Office of Scientific and Technical Information 

(OSTI) 

Report on surveys of irradiation facilities, stable isotope and heavy 

element inventories, radioisotope inventories, Health Physics reports, 

reports of shipments, the first neptunium processing campaign, U-233 

in irradiated thorium, dosimetry service performance testing, particulate 

diffusion near buildings, and the MESODIF-II description and code. 

01/21/2015 37 

DOL - Paragon Mixed waste inventory characteristics, Organic Moderated Reactor 

Experiment (OMRE) evaluations, and transcripts of the National 

Academies of Science committee meetings on remediation of buried 

waste. 

01/23/2012 5 

East Tennessee Technology Park (ETTP) Records Center Report confirming the routing of some recycled uranium through the 

ICPP. 

06/05/2014 1 

Federal Records Center (FRC) - Atlanta The 1986 Effluent Information System (EIS)/Onsite Discharge 

Information System (ODIS) executive summary. 

03/16/2004 1 

Federal Records Center (FRC) - Boston Personnel termination exposure summaries. 04/27/2012 1 

Federal Records Center (FRC) - Chicago Personnel monitoring records. 09/29/2008 2 

Federal Records Center (FRC) - Denver Mixed waste streams, a review of criticality accidents, a 1995 DOE 

occupational exposure report, Photodosimetry Evaluation Book Volume 

IV, and the transfer of the tritium plasma experiment glovebox to INL. 

01/31/2012 5 

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Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Federal Records Center (FRC) - Kansas City Argonne National Laboratory (ANL) fuel cycle and waste management 

field work proposals and agreements. 

08/15/2008 1 

Federal Records Center (FRC) - San Bruno Operations reports, equipment development reports, Health Physics 

reports, and bioassay procedures. 

08/03/2012 16 

General Atomics An SL-1 accident report, 1960-1964 radiation injury claims, material 

shipping and status reports, and GAMAS Trailer exposures at INL. 

01/10/2006 5 

General Electric (GE) - Evendale The history of GE at INL and the disposition of records. 11/12/2010 1 

Hagley Museum & Library Mention of INL as a major liquid waste producer and a trip report to 

ANL. 

09/28/2010 2 

Hanford AEC operational accidents and radiation exposures 1943-1970, an SL-1 

report, Hanford operational reports, neptunium production and 

recovery, material balance and transaction reports, and DOE comments 

on the Environmental Protection Agency's proposed hazardous waste 

management system. 

01/17/2013 29 

Hanford / S. Cohen & Associates (SC&A) A routine fecal sampling program for plutonium workers. 08/05/2004 1 

[Name Redacted] A dosimetry system intercomparison report. 08/13/2003 1 

Idaho National Laboratory / SC&A Incident reports, Initial Engine Test reports, stack monitoring 

description and data, and 1985 TLD and air monitoring data for the 

Radioactive Waste Management complex (RWMC). 

06/05/2012 14 

Interlibrary Loan A 1973 criticality short course, 1958 uranium ore processing, 

environmental reports, INL participation in a tritium plasma 

experiment, and ICPP technical progress reports. 

01/13/2015 28 

Internet - Defense Technical Information Center (DTIC) Annual reports to Congress on activities relating to the Defense Nuclear 

Facilities Safety Board (DNFSB), plutonium stabilization, waste 

treatment reports, irradiation reports, and radiochemical constituents of 

aqueous effluents. This database has not been searched in its entirety. 

OPEN 18 

Internet - Defense Technical Information Center (DTIC) / 

SC&A 

A 1995 site environmental report. 01/09/2012 1 

Internet - DOE The standard of good radiological protection practices in plutonium 

facilities and experimental data for release fractions from nonreactor 

facilities. 

12/04/2008 2 

Internet - DOE Comprehensive Epidemiologic Data 

Resource (CEDR) 

This database has not been searched in its entirety. OPEN 0 

Internet - DOE Environmental Management Linking Legacies Chapter 3: Wastes. 10/28/2007 1 

248 of 258 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Internet - DOE Hanford DDRS Hanford monthly reports referencing INL, criticality considerations and 

loads forecasts for ICPP, material shipments, and plutonium for the 

Zero Power Physics Reactor (ZPPR). 

06/06/2013 40 

Internet - DOE Legacy Management INL decontamination and decommissioning technology logic diagrams 

and a DOE mixed waste report. 

09/19/2014 3 

Internet - DOE Legacy Management Considered Sites Materials packaging and shipping reports and a mixed waste report. 12/29/2014 3 

Internet - DOE National Nuclear Security Administration 

(NNSA) - Nevada Site Office 

No relevant data identified. 09/19/2012 0 

Internet - DOE Oak Ridge Operations Office An interim action report for fuel and flush salt removal from the Oak 

Ridge National Laboratory (ORNL) molten salt reactor experiment. 

01/14/2014 1 

Internet - DOE OpenNet Incident reports, interviews, environmental branch reports, aircraft 

nuclear propulsion reports, SNAPTRAN reports, Health Physics 

reports, iodine skin absorption reports and consent documents, SPERT 

reports, and 1961 site radiation protection standards. 

12/29/2014 86 

Internet - DOE OpenNet / Hanford A human radiation studies interview. 11/26/2007 1 

Internet - DOE OpenNet / NIOSH A 1960 annual report to Congress. 01/11/2008 1 

Internet - DOE OSTI The Tiger Team Assessment of INL, 1965 and 1969 waste disposal 

data, and a nuclear waste management report. 

11/11/2013 6 

Internet - DOE OSTI Energy Citations Characterization of transuranic waste, nuclear fuel development and 

research reports, fuel transfers, facility descriptions, external dosimetry 

reports, environmental surveys and assessments, mixed waste 

assessments, effluent monitoring, ICPP upgrade evaluations, Nevada 

Test Site shipment reports, thorium utilization report, reactor core 

experiments, advanced neutron source report, reactor operations 

procedures, nuclear air cleaning conference proceedings, and detection 

of highly-enriched uranium. 

05/07/2013 100 

Internet - DOE OSTI Energy Citations / SC&A A 1966 controlled environmental radioiodine test. 04/29/2010 1 

249 of 258 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Internet - DOE OSTI Information Bridge Integrated spent fuel and radioactive waste inventories, environmental 

reports to Congress, interviews, waste generation reports, reactor fuel 

reports, annual DOE radioisotope customer and shipment reports, 

environmental reports, ZPPR progress reports, continuous detection of 

airborne Pu-239, spent fuel treatment, decontamination and 

decommissioning reports, air emission reports, radiological control 

performance indicator reports, transuranic control reports, ICPP 

research reports, properties of Rocky Flats vitrified transuranic waste, 

disposal of high level waste at the ICPP, handling ICPP high level 

waste, EBR-II reports, U-233 recovery, neptunium processing, and low 

level waste transportation. 

08/13/2013 498 

Internet - DOE OSTI Information Bridge / SC&A A 1954 BORAX experiment Health Physics report. 07/06/2007 1 

Internet - DOE OSTI SciTech Connect The INEL historical dose evaluation-reconstruction of airborne releases, 

chemical processing technology periodic reports, technical progress 

reports, analytical branch reports, safety analyses reports, Advanced 

Test Reactor (ATR) radiological monitoring results, ATR waste reports, 

instrument branch reports, analyses of irradiated materials, ICPP waste 

handling and management reports, fuels and materials reports, metals 

and ceramics reports, environmental reports, air emissions reports, and 

mixed waste processing reports. 

01/13/2015 427 

Internet - Energy Employees Claimant Assistance Project 

(EECAP) 

Compendium of federal and state radioactive materials transportation 

laws and regulations and Arco, Idaho soil samples. This database has 

not been searched in its entirety. 

OPEN 2 

250 of 258 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Internet - Google Environmental reports, ABRWH meeting minutes, waste reports, 

nuclear nonproliferation reports, U-233 storage safety, DOE 

radioisotope customers, DOE occupational exposure reports, INL 

history and photographs, waste shipments, reports to Congress, 

environmental impact statements, a 2004 union agreement, a 

comprehensive RWMC waste inventory and projection, Aircraft 

Nuclear Propulsion reports, SL-1 reports, radionuclide releases, 

environmental behavior of radionuclides, air sampling, remediation 

conceptual designs, EPA records of decision, facility descriptions, risk 

assessments, INL oversight reports, ICPP operational reports, 

instrument and equipment instructions and reports, Rocky Flats 

transuranic waste reports, estimating releases of various radionuclides, 

radionuclide releases, and planning for the long term storage of high 

level ICPP waste. 

09/22/2014 785 

Internet - Hathitrust ICPP technical progress reports. 12/02/2014 4 

Internet - Health Physics Journal Health Physics aspects of the SL-1 accident, environmental reports 

including intakes by indigenous animal species, iodine monitoring, the 

rem response of the INL personnel dosimeter to photons, and 

application of ALARA to the high level waste tank closure project. This 

database has not been searched in its entirety. 

OPEN 15 

Internet - Journal of Occupational and Environmental 

Hygiene 

This database has not been searched in its entirety. OPEN 0 

Internet - National Academies Press (NAP) Waste treatment and disposition, plutonium treatment and disposition, 

radionuclide releases, review of DOE's cleanup roadmap, and cancer 

risks to adjacent populations. 

12/29/2014 11 

Internet - National Institute for Occupational Safety and 

Health (NIOSH) 

Residual beryllium contamination reports, a review of Initial Engine 

Test source terms, and the epidemiologic study of mortality and cancer 

risk at INL 

11/02/2011 5 

Internet - NRC Agencywide Document Access and 

Management (ADAMS) 

A FOIA response for Westinghouse Electric, waste reports, spent 

nuclear fuel management, plutonium and highly-enriched uranium 

management, weapons-usable material disposition environmental 

impact statement and studies, LOFT experiment reports, studies on 

environmental transport, environmental reports, NRC reviews of INL 

remediation plans, facility safety assessments, and accident scenario 

assessments. 

12/29/2014 205 

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Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Internet - Oak Ridge National Laboratory (ORNL) ORNL division progress reports referencing INL and summary of 

transuranic waste assay methods used by INL. 

04/08/2013 102 

Internet - University of Hawaii, Manoa A 1965 aerial survey of the National Reactor Testing Station. 11/08/2007 1 

Internet - University of North Texas A reactor engineering division quarterly report, extraction study reports, 

summary of the SPERT reactor facilities, Army gas-cooled reactor 

periodic reports, hazards summary reports, a SNAPTRAN safety 

analysis, and the microfilm system for engineering drawings. 

11/11/2014 29 

Internet - US Army Corps of Engineers (USACE) No relevant data identified. 12/29/2014 0 

Internet - US Environmental Protection Agency NEPIS Superfund records of decision, environmental I-129, and 1987 mixed 

energy waste studies. 

12/29/2014 15 

Internet - US Transuranium and Uranium Registries A mortality and longevity report on workers exposed to plutonium. 12/29/2014 1 

Kansas City Plant Radiation exposure records and personal dosimeter procedures. 10/09/2013 5 

Lawrence Livermore National Laboratory Request for approval to ship Tory II-A fuel elements. 08/02/2014 1 

Los Alamos National Laboratory Photodosimetry Evaluation Book "Bible" Volume VIII Procedures 

1996-2001 (LANL 2001b) and radioactive waste disposal issues. 

11/29/2006 2 

Los Alamos National Laboratory - LAHDRA Waste projections, characteristics of mixed waste streams, and storage 

and disposition of weapons-usable material. 

12/13/2007 6 

Missouri Department of Natural Resources Plutonium Working Group reports. 10/01/2008 3 

Mound Museum Bismuth reports, weapons production, waste management, and Mound 

publications referencing INL. 

02/01/2012 13 

National Archives and Records Administration (NARA) ­

Atlanta 

A Materials Testing Reactor report, accident report summaries, a 

National Reactor Testing Station information packet, the DOE indoor 

radon study, and directory of consultants. 

05/23/2008 12 

National Archives and Records Administration (NARA) ­

Atlanta / SC&A 

SL-1 reports. 09/26/2003 2 

National Archives and Records Administration (NARA) ­

College Park 

Aircraft nuclear Propulsion reports, a Rover/Tory fuel disposal report, 

and the 1959 report on the future role of AEC laboratories. 

08/17/2010 4 

National Archives and Records Administration (NARA) ­

Seattle 

Incident reports, ETR, TAN, CFA, MTR, SPERT Health Physics 

reports, radiation exposure reports, shipment records, radioactive waste 

reports, radiological survey records, Safety Section reports, Burial 

Ground reports, and Industrial Hygiene reports. 

12/18/2014 205 

National Archives and Records Administration (NARA) ­

Seattle / SC&A 

1954 and 1955 environmental records. 11/15/2006 2 

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Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

National Institute for Occupational Safety and Health 

(NIOSH) 

Internal dose assessment and bioassay manuals and reports, 

environmental transport reports, reports to Congress, the AEC history, 

interviews, worker outreach documents, radiological surveys, effluent 

releases, and plutonium bioassay and contamination reports. 

12/31/2014 94 

National Institute for Occupational Safety and Health 

(NIOSH) / SC&A 

Highly Enriched Uranium Working Group reports, a recycled uranium 

report, an INL former worker medical surveillance program report, and 

a study on aerosols generated by metal cutting techniques. 

10/20/2006 7 

National Nuclear Security Administration (NNSA), 

Albuquerque Office 

A 1996 external dosimetry report. 09/03/2008 1 

National Technical Information Service (NTIS) Spent fuel storage casks tests and analyses. 02/19/2010 2 

New York State Archives Waste disposal and 1952 periodic reports. 03/19/2012 2 

Nevada Test Site The Tory II-C fact sheet and a Nevada environmental impact statement. 01/12/2003 2 

Nuclear Regulatory Commission Non-Public Holdings Redacted radiation exposure reports and a blowdown rests report. 06/04/2012 3 

Nuclear Regulatory Commission Public Document Room Inspection reports, management of transuranic waste, material transfer 

reports, and the INEEL emergency plan and RCRA contingency plan. 

12/17/2014 21 

Oak Ridge Library for Dose Reconstruction Design of an iodine removal unit from the ICPP waste gas stream, 

waste treatment reports, MTR-B project waste solutions, and Oak Ridge 

National Laboratory reports referencing INL. 

05/10/2011 14 

Oak Ridge National Laboratory (ORNL) Eleventh DOE workshop on personnel neutron dosimetry, the transfer 

of U-233 to INL, request for an irradiation in the MTR, and ORNL 

periodic reports referencing INL. 

03/20/2014 55 

Ohio Department of Health An Aircraft Nuclear Propulsion report, a 1953 Initial Engine Test 

report, and an environmental restoration report. 

11/03/2008 3 

ORAU Team Technical basis documents, technical information bulletins, annual 

DOE radiation exposure reports, effluent monitoring, newspaper 

articles, counting systems calibrations, whole body counting reports, 

documented communications, instrumentation descriptions and 

manuals, and DOE adoption of 1990 ICRP neutron weighting factors. 

11/25/2014 106 

Rocky Flats Environmental Technology Site (RFETS) An assessment of the flammability and explosion potential of defense 

transuranic waste. 

05/17/2006 1 

SAIC Radiation exposure summaries. 09/02/2004 8 

Sandia National Laboratory - NM Radiation exposure requests and Sandia Engineering Reactor Health 

Physics logs. 

01/17/2012 5 

Santa Susana Field Laboratory Organic Moderated Reactor Experiment reports. 08/10/2009 26 

Savannah River Site (SRS) Dosimetry visitor cards and SRS periodic reports referencing INL. 03/19/2012 35 

253 of 258 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

S. Cohen & Associates (SC&A) Environmental reports, SNAPTRAN reports, incident reports, the 

RWMC history, an SL-1 report, inventory and manufacturing 

statements, ICPP production reports, and a documented communication. 

04/07/2011 101 

SC&A / ANL-E A report on planned processing of irradiated plutonium fuel assemblies, 

the production of transplutonium elements, and a description of ANL 

critical facilities. 

06/24/2010 3 

SC&A / INL Environmental monitoring reports, incident reports, periodic facility 

reports, history of buried transuranic waste at INL, construction staff 

meeting notes, ICPP process reports, remedial investigation reports and 

work plans, SL-1 reports, ground water reports, radiochemistry reports, 

pollution control, effluent releases, Initial Engine Test reports, 

decontamination and decommissioning reports, material inventories, lab 

notebooks, Health Physics monthly reports, and source term quality 

control tasks. 

06/24/2010 1,980 

SC&A / INL / Seattle NARA 1956 monthly radioactive waste reports. 06/24/2010 1 

SC&A / Internet - OpenNet The first 50 years plutonium history. 10/28/2014 1 

SC&A / Internet - DOE OSTI Information Bridge The 1992 review of the Radiological Environmental Surveillance 

Program. 

08/06/2010 1 

SC&A / Internet - Google A 1959 waste calcination off-gas study. 09/02/2014 1 

SC&A / Internet - University of North Texas Design and hazards summary report on the Boiling Reactor Experiment 

V (Borax V). 

10/13/2014 1 

SC&A / Pinellas Plant A 1993 waste generation and minimization report. 06/24/2010 1 

SC&A / West Valley Demonstration Project Quality control of West Valley TLDs at DOE-ID, procedure for 

releasing exposure histories, procedures for the records interface with 

DOE-ID, special dose evaluations, and preparation of samples for 

shipment to INL. 

05/06/2011 7 

Senator John Heinz History Center A history of Westinghouse in atomic power. 12/20/2007 1 

Southern Illinois University, Edwardsville, IL AEC construction cost differentials and the SC&A review of the 

Mallinckrodt Site Profile which mentions the similarity between the 

Mallinckrodt and INL external dosimeter. 

10/18/2008 2 

University of Iowa A records listing and the DOE Environmental Site Profile for INL. 07/11/2003 2 

University of Tennessee Library The report of Plutonium Task Force 7 and an SL-1 report. 10/03/2011 2 

254 of 258 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-1: Data Capture Synopsis for Idaho National Lab 

Data Capture Information General Description of Documents Captured Date 

Completed 

Uploaded 

To SRDB 

Unknown An SL-1 report, safety analyses and hazard summaries, reactor 

procedures, effluent releases, environmental reports, radiation exposure 

reports, in-vivo reports, internal dosimetry incident reports, external 

dosimetry technology, technical bases, procedure reports, skin 

contaminations, and conferences on the peaceful uses of atomic energy. 

06/28/2010 233 

Unknown / INL Fluidized bed calcination of aluminum nitrate wastes, 1971 Health 

Services Laboratory report, and 1979 environmental monitoring data. 

05/10/2012 3 

Unknown / SC&A Environmental reports, the EBR-1 meltdown, and the SPERT-1 

destructive test report. 

10/24/2003 42 

Unknown / SC&A / INL A 1959 Health and Safety Division report. 10/24/2003 1 

Washington University Library A lithium cooled reactor report. 04/23/2007 1 

West Valley Demonstration Project A 1981 investigation of a radiation exposure at the MTR Plug Storage 

Area. 

08/02/2006 1 

Wyeth Insurance A 1951 newsletter report on American Cyanamid winning the ICPP 

operating contract. 

02/29/2008 1 

Y-12 / SC&A A report on plutonium contamination on recycled material from INL. 07/28/2010 1 

TOTAL 9,350 

Table A1-2: Databases Searched for Idaho National Laboratory 

Database/Source Keywords / Phrases Hits Selected 

NOTE: Database search terms employed for each of the databases listed below 

are available in the Excel file Called “Idaho National Laboratory Rev 00, (83.13) 02-06-15 

Defense Technical Information Center (DTIC) 

https://www.dtic.mil/ 

See Note above 

OPEN 

DOE CEDR 

https://www.orau.gov/cedr 

See Note above 

OPEN 

255 of 258 
 

 

 

 

 

 

 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-2: Databases Searched for Idaho National Laboratory 

Database/Source Keywords / Phrases Hits Selected 

DOE Hanford DDRS 

http://www2.hanford.gov/declass/ 

COMPLETED 09/18/2012 

See Note above 32 12 

DOE Legacy Management Considered Sites 

http://www.lm.doe.gov/considered_Sites/ 

COMPLETED 12/29/2014 

See Note above 2,795 4 

DOE NNSA - Nevada Site Office 

www.nv.doe.gov/main/search.htm 

COMPLETED 09/19/2012 

See Note above 0 0 

DOE OpenNet 

http://www.osti.gov/opennet/advancedsearch.jsp 

COMPLETED 12/29/2014 

See Note above 7,641 75 

DOE OSTI Energy Citations 

http://www.osti.gov/energycitations/ 

COMPLETED 09/19/2012 

See Note above 47,027 39 

DOE OSTI Information Bridge 

http://www.osti.gov/bridge/advancedsearch.jsp 

COMPLETED 09/19/2012 

See Note above 41,900 221 

DOE OSTI SciTech Connect 

http://www.osti.gov/scitech 

COMPLETED 12/29/2014 

See Note above 2,687,639 634 

Energy Employees Claimant Assistance Project 

(EECAP) 

http://www.eecap.org 

See Note above 

OPEN 

Google 

http://www.google.com 

COMPLETED 09/22/2014 

See Note above 146,556,657 757 

HP Journal 

http://journals.lww.com/health-physics/pages/default.aspx 

See Note above OPEN 

Journal of Occupational and Environmental Health 

http://www.ijoeh.com/index.php/ijoeh 

See Note above OPEN 

National Academies Press 

http://www.nap.edu/ 

COMPLETED 12/29/2014 

See Note above 56,359 7 

256 of 258 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-2: Databases Searched for Idaho National Laboratory 

Database/Source Keywords / Phrases Hits Selected 

NEPIS 

http://nepis.epa.gov/ 

COMPLETED 12/29/2014 

See Note above 40,154 31 

NRC ADAMS Reading Room 

http://www.nrc.gov/reading-rm/adams/web-based.html 

COMPLETED 12/29/2014 

See Note above 30,586 125 

United States Army Corps of Engineers (USACE) 

http://www.usace.army.mil/ 

COMPLETED 12/29/2014 

See Note above 10 0 

U.S. Transuranium & Uranium Registries 

http://www.ustur.wsu.edu/ 

COMPLETED 12/29/2014 

See Note above 930 1 

Table A1-3: Interlibrary Loan Documents Requested for Idaho National Laboratory 

Document Number Document Title Date Requested Date Received 

NA 

Ref ID: 32032 

A Whole-Body Counter with Rotating Detectors from Health Physics 

Journal Vol. 16:709-717 

03/20/2013 03/20/2013 

NA 

Ref ID: 32031 

A Portable Whole-Body Counter from Health Physics Journal Vol 

13:141-148 

03/20/2013 03/20/2013 

IDO-14453 

Ref ID: 139094 

Technical Progress Report for April through June 1958: Idaho Chemical 

Processing Plant 

11/11/2014 01/13/2015 

IDO-14457 

Ref ID: 139096 

Technical Progress Report for July through September 1958: Idaho 

Chemical Processing Plant 

11/11/2014 01/13/2015 

IDO-14467 

Ref ID: 139097 

Technical Progress Report for October through December 1958: Idaho 

Chemical Processing Plant 

11/11/2014 01/13/2015 

IDO-14471 

Ref ID: 139095 

Technical Progress Report for January through March 1959: Idaho 

Chemical Processing Plant 

11/11/2014 01/13/2015 

IDO-14494 

Ref ID: 139093 

Technical Progress Report for April through June 1959: Idaho Chemical 

Processing Plant 

11/11/2014 01/13/2015 

IDO-14526 

Ref ID: 139092 

Idaho Chemical Processing Plant Technical Progress Report: 

Radioactive Waste Disposal Projects, October-December 1959 

11/11/2014 01/13/2015 

257 of 258 
 

 

 

 

SEC-00219 03-12-15 Idaho National Laboratory 

Table A1-3: Interlibrary Loan Documents Requested for Idaho National Laboratory 

Document Number Document Title Date Requested Date Received 

IDO-14530 

Ref ID: 139090 

Idaho Chemical Processing Plant Technical Progress Report: 

Radioactive Waste Disposal Projects, January-March 1960 

11/11/2014 01/13/2015 

IDO-14514 

Ref ID: 139091 

Idaho Chemical Processing Plant Technical Progress Report, 

Radioactive Waste Disposal Projects, July-September 1959 

11/11/2014 01/13/2015 

258 of 258